Elevated-Temperature Mechanical Properties of an Advanced-Type 316 Stainless Steel1

Brinkman, C.R.

doi:10.1115/1.1343911

Cited by 34 publications

(20 citation statements)

References 4 publications

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“…Fatigue lives of 316FR and 316 are almost the same at 500-600°C. However, when hold time was applied to the cyclic loading, 316FR showed significantly improved creep-fatigue resistance over 316SS as shown in Figure 4.51 (Brinkman 2001). The improvements were more pronounced at longer hold time and at lower strain ranges.…”

Section: Creep and Creep-fatigue Properties And High Temperature Desimentioning

confidence: 95%

“…316LN-IG has higher allowable stresses than typical 316LN due to an optimal combination of the alloying elements such as carbon, nitrogen, nickel, chromium, manganese and molybdenum, with a tight specification of their allowable composition range. A comprehensive database has been established on 316LN-IG, including heat-to-heat variations, the effect of product size, fracture toughness, effects of neutron irradiation, etc., and the data can be found in the ITER Materials Properties Handbook ( The chemical compositions of 316LN, 316LN-IG, and 316FR stainless steels are given in Table 4-11 (Asayama 2001, Tavassoli 1995, Brinkman 2001. Significant data base of tensile, creep, fatigue and creep-fatigue properties of 316FR and 316LN base metal and weld metal can be found in the RCC-MR, ITER MPH and open literature.…”

Section: Fr and 316ln Stainless Steelsmentioning

confidence: 99%

“…stainless steels tested at 500, 550 and 600°C (Brinkman 2001). The data set consist of data for several heats and various product forms (e.g.…”

Section: Creep and Creep-fatigue Properties And High Temperature Desimentioning

confidence: 99%

“…Fatigue data of 316FR are shown in Figure 4.50, and compared with conventional 316SS (Brinkman 2001). Fatigue lives of 316FR and 316 are almost the same at 500-600°C.…”

Section: Creep and Creep-fatigue Properties And High Temperature Desimentioning

confidence: 99%

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Code qualification of structural materials for AFCI advanced recycling reactors.

Natesan¹,

Li²,

Nanstad³

et al. 2012

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This report summarizes the further findings from the assessments of current status and future needs in code qualification and licensing of reference structural materials and new advanced alloys for advanced recycling reactors (ARRs) in support of Advanced Fuel Cycle Initiative (AFCI). 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 AFCI Reactor Campaign. The report is the second deliverable in FY08 (M505011401) under the work package "Advanced Materials Code Qualification".The overall objective of the Advanced Materials Code Qualification project is to evaluate key requirements for the ASME Code qualification and the Nuclear Regulatory Commission (NRC) approval of structural materials in support of the design and licensing of the ARR. 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. Code qualification and licensing of advanced materials are prominent needs for developing and implementing advanced sodium reactor technologies. Nuclear structural component design in the U.S. must comply with the ASME Boiler and Pressure Vessel Code Section III (Rules for Construction of Nuclear Facility Components) and the NRC grants the operational license. As the ARR will operate at higher temperatures than the current light water reactors (LWRs), the design of elevated-temperature components must comply with ASME Subsection NH (Class 1 Components in Elevated Temperature Service). However, the NRC has not approved the use of Subsection NH for reactor components, and this puts additional burdens on materials qualification of the ARR. In the past licensing review for the Clinch River Breeder Reactor Project (CRBRP) and the Power Reactor Innovative Small Module (PRISM), the NRC/Advisory Committee on Reactor Safeguards (ACRS) raised numerous safety-related issues regarding elevated-temperature structural integrity criteria. Most of these issues remained unresolved today. These critical licensing reviews provide a basis for the evaluation of underlying technical issues for future advanced sodium-cooled reactors.Major materials performance issues and high temperature design methodology issues pertinent to the ARR are addressed in the report. The report is organized as follows: the ARR reference design concepts proposed by the Argonne National Laboratory and four industrial consortia were reviewed first, followed by a summary of the major code qualification and licensing issues for the ARR structural materials. The available database is presented for the ASME Code-qualified structural alloys (e.g. 304, 316 stainless steels, 2.25Cr-1Mo, and mod.9Cr-1Mo), including physical properties, tensile properties, impact properties and fracture toughness, creep, fatigue, creep-fatigue in...

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Section: Creep and Creep-fatigue Properties And High Temperature Desimentioning

confidence: 95%

Section: Fr and 316ln Stainless Steelsmentioning

confidence: 99%

“…stainless steels tested at 500, 550 and 600°C (Brinkman 2001). The data set consist of data for several heats and various product forms (e.g.…”

Section: Creep and Creep-fatigue Properties And High Temperature Desimentioning

confidence: 99%

“…Fatigue data of 316FR are shown in Figure 4.50, and compared with conventional 316SS (Brinkman 2001). Fatigue lives of 316FR and 316 are almost the same at 500-600°C.…”

Section: Creep and Creep-fatigue Properties And High Temperature Desimentioning

confidence: 99%

See 2 more Smart Citations

Code qualification of structural materials for AFCI advanced recycling reactors.

Natesan¹,

Li²,

Nanstad³

et al. 2012

View full text Add to dashboard Cite

show abstract

“…For an illustration, a plot of the creep rupture data at different test temperatures is shown in Fig. 5.19 for 316 stainless steels, from Brinkman (2001). It is typical that a majority of data in a creep rupture database consists of rupture times that are less than 50,000 to 60,000 hours.…”

Section: Asme Design Limitsmentioning

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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The Mechanical Behavior of a 25Cr Super Duplex Stainless Steel at Elevated Temperature

Lasebikan

Akisanya

Deans

2012

J. of Materi Eng and Perform

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Super duplex stainless steel (SDSS) is a candidate material for production tubing in oil and gas wells and subsea pipelines used to transport corrosive hydrocarbon fluids. The suitability of this material for high temperature applications is examined in this paper. The uniaxial tensile properties are determined for a 25Cr super duplex stainless steel over a range of temperature relevant to high pressure-high temperature oil and gas wells. It is shown that there is a significant effect of temperature on the uniaxial tensile properties. Elevated temperature was shown to reduce the Young's modulus and increase the strain hardening index; temperature effects on these two parameters are usually neglected in the design of subsea pipelines and oil well tubulars and this could lead to wrong predictions of the collapse pressure. The manufacturing process of the super duplex tubular did not lead to significant anisotropy in the hardness and the ultimate tensile and uniaxial yield strengths.

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Elevated-Temperature Mechanical Properties of an Advanced-Type 316 Stainless Steel1

Cited by 34 publications

References 4 publications

Code qualification of structural materials for AFCI advanced recycling reactors.

Code qualification of structural materials for AFCI advanced recycling reactors.

Resolution of qualification issues for existing structural materials.

The Mechanical Behavior of a 25Cr Super Duplex Stainless Steel at Elevated Temperature

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