Effect of sodium environment on the low cycle fatigue properties of modified 9Cr–1Mo ferritic martensitic steel

Kannan, Rangasayee; Sandhya, R.; Ganesan, V.; Valsan, M.; Rao, K. Bhanu Sankara

doi:10.1016/j.jnucmat.2008.11.036

Cited by 30 publications

(9 citation statements)

References 15 publications

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“…The calculated cyclic stress curves well capture the trends of cyclic softening response of G91 steel: the peak stress decreases with increasing number of cycles, and cyclic softening effect is stronger at a higher total strain range. The calculated cyclic softening curves were compared with the literature data [Kannan et al 2009, Shankar et al 2006. A reasonable agreement was achieved between the experimental data and model predictions.…”

Section: Cyclic Softening Modelmentioning

confidence: 55%

Final report on improved creep-fatigue models on advanced materials for SFR applications.

Li¹,

Soppet²,

Majumdar³

et al. 2012

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This report provides an update on the materials performance criteria and methodology relevant to sodium-cooled fast reactors (SFRs). The report is the second deliverable (Level 2) in FY11 (M2A11AN040303) under the work package A-11AN040303 "Materials Performance Criteria and Methodology" as part of the Advanced Structural Materials Program for the Advanced Reactor Concepts.The overall objective of the Materials Performance Criteria and Methodology work project is to evaluate the key requirements for the ASME Code qualification and the Nuclear Regulatory Commission (NRC) approval of advanced structural materials in support of the design and performance of sodium-cooled fast reactors. Advanced materials are a critical element in the development of fast reactor technologies. Enhanced materials performance not only improves safety margins and provides design flexibility, but also is essential for the economics of future advanced fast reactors. Qualification and licensing of advanced materials are prominent needs for the development and implementation of advanced fast reactor technologies. Nuclear structural component designs in the U.S. comply with the ASME Boiler and Pressure Vessel (B&PV) Code Section III (Rules for Construction of Nuclear Facility Components), and the NRC grants licensing. As the SFRs 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). A number of technical issues relevant to materials performance criteria and high temperature design methodology in the SFR were identified and presented in earlier reports. A viable approach to resolve these issues and the R&D priority were also recommended. The development of mechanistically based creep-fatigue interaction models for life prediction and reliable data extrapolation was chosen to be the central focus in near-term efforts.Our current focus is on the creep-fatigue damage issue in high-strength ferritic/martensitic steels such as mod.9Cr-1Mo (G91) and NF616 (G92) steels. The current ASME creep-fatigue design rule puts severe limits of fatigue and creep loads for G91 steel, the lead structural material for fast reactors. High-strength ferritic/martensitic steels behave fundamentally differently from austenitic stainless steels, for which the current ASME creep-fatigue design rules were developed. The unique deformation and damage characteristics in G91 steel, e.g. cyclic softening, degradation of creep and rupture strength during cyclic service, demands a new creepfatigue design procedure that explicitly accounts for the material's unique creep-fatigue behavior. G92 steel, a variant of G91 steel, is the lead candidate in the Advanced Alloy Development program. There is currently no creep-fatigue design rule available for this advanced alloy.To support the predictive model development and resolve the over-conservative issue with the ASME design rule for G91 steel, we reco...

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Section: Cyclic Softening Modelmentioning

confidence: 55%

Final report on improved creep-fatigue models on advanced materials for SFR applications.

Li¹,

Soppet²,

Majumdar³

et al. 2012

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“…This difference is likely due to heat-to-heat variations as well as in differences in material strength resulting from heat treatments of G91 steel. A tensile strength of 380 MPa was used to calculate the cyclic softening curve for the steel studied by Kannan et al [2009] and a tensile strength of 345 MPa was used for the calculation of cyclic softening curve for the G91 steel investigated by Shankar et al [2006]. A good agreement was achieved in both cases between the experimental data and model predictions.…”

Section: Cyclic Softening Model For G91 Steelmentioning

confidence: 89%

“…The calculated cyclic softening curves were compared with the experimental data reported in the literature [Kannan et al 2009, Shankar et al 2006. The comparisons are shown in Fig.…”

Section: Cyclic Softening Model For G91 Steelmentioning

confidence: 99%

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Status report on improved understanding of creep-fatigue damage in advanced materials.

Li¹,

Majumdar²,

Soppet³

et al. 2012

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This report provides an update on the materials performance criteria and methodology relevant to liquid metal reactors (LMRs), in particular, sodium cooled fast reactors. The report is the first deliverable (Level 3) in FY11 (M3A11AN04030303) under the work package A-11AN040303 "Materials Performance Criteria and Methodology" as part of Advanced Structural Materials Program for the Advanced Reactor Concepts.The overall objective of the Advanced Materials Performance Criteria and Methodology project is to evaluate the key requirements for the ASME Code qualification and the Nuclear Regulatory Commission (NRC) approval of advanced structural materials in support of the design and licensing of the liquid metal fast reactors. Advanced materials are a critical element in the development of fast reactor technologies. Enhanced materials performance not only improves safety margins and provides design flexibility but also is essential for the economics of future advanced fast reactors. Qualification and licensing of advanced materials are prominent needs for the development and implementation of advanced fast reactor technologies. Nuclear structural component designs in the U.S. comply with the ASME Boiler and Pressure Vessel (B&PV) 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). A number of technical issues relevant to materials performance criteria and high temperature design methodology in the LMR were identified and presented in earlier reports [Natesan et al. 2008[Natesan et al. , 2009. A viable approach to resolve these issues and the R&D priority were also recommended. The development of mechanistically based creep-fatigue interaction models for life prediction and reliable data extrapolation was chosen to be the central focus in near-term efforts.Our current efforts focus on the creep-fatigue damage issue in high-strength ferriticmartensitic steels for two primary reasons. First, the current ASME design rule of bilinear damage summation puts severe limits of fatigue and creep loads for mod.9Cr-1Mo (G91) ferriticmartensitic steel, the lead structural material for fast reactors; secondly, the ferritic-martensitic steels behave fundamentally differently from austenitic stainless steels, for which the current ASME creep-fatigue design rules were developed. The unique deformation and damage characteristics in G91 steel, e.g. cyclic softening, degradation of creep and rupture strength during cyclic service, demands a new creep-fatigue design procedure that explicitly accounts for the material's unique creep-fatigue behavior. To support the development of predictive models and to resolve the over-conservative issue with the ASME design rule for G91 steel, we recovered stress relaxation data from thirteen creep-fatigue te...

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“…Sodium exposure has little effect on fatigue life of G91. Kannan et al (2009) reported the fatigue data of mod.9Cr-1Mo steel tested in flowing sodium. Fatigue tests were conducted under fully reversed, total axial strain control at 550 and 600°C.…”

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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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Effect of sodium environment on the low cycle fatigue properties of modified 9Cr–1Mo ferritic martensitic steel

Cited by 30 publications

References 15 publications

Final report on improved creep-fatigue models on advanced materials for SFR applications.

Final report on improved creep-fatigue models on advanced materials for SFR applications.

Status report on improved understanding of creep-fatigue damage in advanced materials.

Resolution of qualification issues for existing structural materials.

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