Microstructural Effects on Creep-Fatigue Life of Alloy 709

McMurtrey, Michael D; Carroll, Laura; Wright, J.K.

doi:10.2172/1404723

Cited by 2 publications

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“…In this study, we investigate the interactions between dislocations and displacement cascades using MD simulations in Fe-20Cr-25Ni alloy. The Fe-20Cr-25Ni is the base system of Alloy 709 (47,48), which is currently considered as a candidate of structural materials for next generation nuclear reactors (49).…”

Section: Iv-1 Interaction Between Displacement Cascades and Dislocatimentioning

confidence: 99%

Integrated Computational Study of Radiation Damage Effects in Grade 92 Steel and Alloy 709

Tan

Sridharan

et al. 2018

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Radiation resistance of austenitic alloy 709 (named as A709) and ferritic-martensitic steel Grade 92 (G92) were investigated in this project using 316H and Grade 91 (G91) as references, respectively. In the meantime, a model 709 (M709) alloy (Fe-20Cr-24Ni) was prepared to study the matrix response to irradiation. Samples machined from the tab section of three creep-ruptured A709 samples, i.e., 17,106 h at 600 °C, 12,014 h at 650 °C, and 16,448 h at 700 °C, were included in this work to evaluate radiation effect on the long-term "thermally aged" samples. Primarily Fe 2+ ion irradiation was conducted at 350, 600 and 670 °C for up to 75 displacements per atom (dpa), together with a few proton irradiation experiments at 380-670 °C for up to 1.5 dpa. Additionally, four neutron-irradiated NF709 samples (3.22-8 dpa at ~430-469 °C) were examined. The samples were examined by transmission electron microscopy (TEM), scanning electron microscopy (SEM) and energy dispersive x-ray spectroscopy (EDS) for microstructures and nanoindentation for radiation-hardening. Unlike the negligible hardening after irradiation at 670 °C, irradiation to nominal 2.5-50 dpa at 350 °C resulted in radiation-hardening of 13.5±5.4 % for A709, which was comparable to that of 13.4±6.2 % for M709 and moderately lower than that of 20.1±5.6 % for 316H. The radiation-hardening induced by the same irradiation condition to the "thermally aged" samples was comparable to A709, which were hardened by 15.0±8.2 %, 13.5±6.0 %, and 10.1±6.1 % for the 600, 650, and 700 °C aged samples, respectively. The radiation-hardening of A709 is significantly lower than that (~94±9 %) of neutron-irradiated NF709 to 8 dpa at ~430 °C. Microstructural characterization indicated dislocation loops, type 1/3<111>, in 316H and A709 after irradiation at 350 °C. The loop size and density in A709 were slightly smaller than that in 316H, e.g., 14.2±0.6 nm with 5.2×10 22 m-3 in A709 vs. 15±1 nm with 8.3×10 22 m-3 in 316H after 50 dpa. However, only line dislocations were developed in the samples irradiated at 600-670 °C. A small amount (~2.8×10 12 m-3) of Nb(CN) precipitates in a size of ~72 nm were observed in the as-received A709, which were reduced to ~1.6×10 12 m-3 with a larger size of ~90 nm after irradiation to 50 dpa at 350 °C. A new type of precipitates CrNbN (Z-phase) in ~10 nm was introduced by the rastered beam irradiation to 50 dpa at 670 °C but not by the defocused beam irradiation to 75 dpa at 600 °C. Additionally, some (CrMo)xCy type precipitates were observed in both the high-temperature irradiated A709 samples, which is believed to be the precursor of M23C6. In contrast, a significant amount of multiple types precipitates was identified in 600-700°C aged A709 samples, e.g., (CrMo)23C6, (CrMo)3Ni2Si, Nb(CN), CrNbN, and Ni/Si-rich particles. Large (CrMo)23C6 and (CrMo)3Ni2Si particles, in ~1 µm, grew along grain boundaries, which covered ~53 % I. BACKGROUND Grade 92 and Alloy 709 are being developed as candidate alloys for the Sodium-cooled Fast Reactor (SFR). Grade 92 h...

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Section: Iv-1 Interaction Between Displacement Cascades and Dislocatimentioning

confidence: 99%

Integrated Computational Study of Radiation Damage Effects in Grade 92 Steel and Alloy 709

Tan

Sridharan

et al. 2018

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Tailoring Microstructure of Austenitic Stainless Steel with Improved Performance for Generation-IV Fast Reactor Application: A Review

Chen

Xie

et al. 2023

Crystals

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Austenitic stainless steels are selected as candidate materials for in-core and out-of-core components of Generation-IV fast reactors due to their excellent operating experience in light-water reactors over several decades. However, the performance of conventional austenitic stainless steels proves to be inadequate through operation feedback in fast reactors. To withstand the demands for material performance exposure to the extreme operating environment of fast reactors, modified austenitic stainless steels for in-core and out-of-core components have been developed from the first-generation 300-series steels. The design of an appropriate microstructure becomes a top priority for improving material performance, and key metallurgical features including δ-ferrite content, grain size and secondary phase precipitation pertinent to austenitic stainless steel are focused on in this paper. δ-ferrite content and grain size are closely correlated with the fabrication program and their effects on mechanical properties, especially creep and fatigue properties are critically assessed. Moreover, the impacts of some major elements including nitrogen, stabilization elements (Nb, Ti, V), phosphorus and boron on secondary phase precipitation behaviors during aging or creep are reviewed in detail. Based on the role of the aforementioned metallurgical features, the recommended specification of nitrogen content, stabilization ratio, phosphorus content, boron content, δ-ferrite content and grain size are put forward to guarantee the best-expected performance, which could provide reactors designers with attractive options to optimize fast reactor systems.

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Microstructural Effects on Creep-Fatigue Life of Alloy 709

Cited by 2 publications

References 1 publication

Integrated Computational Study of Radiation Damage Effects in Grade 92 Steel and Alloy 709

Integrated Computational Study of Radiation Damage Effects in Grade 92 Steel and Alloy 709

Tailoring Microstructure of Austenitic Stainless Steel with Improved Performance for Generation-IV Fast Reactor Application: A Review

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