Scripta Metallurgica et Materialia
The microstructure and tensile properties of mitrogen containing vacuum atomized alloy 690
References (14)
- et al.
J. Nuc. Mat.
(1989) - et al.
Scripta Met.
(1989) - et al.
Met. Trans. A
(1990) - et al.
Met. Trans. A
(1989) - et al.
Boshoku Gijutsu
(1979) - et al.
Corrosion
(1990) - et al.
Acta Met.
(1988)
Cited by (23)
Crack growth rate evaluation of alloys 690/152 by numerical simulation of extracted CT specimens
2019, Nuclear Engineering and TechnologyCitation Excerpt :Lots of studies on nickel-based alloys have been carried out by regulatory authorities, research institutes and academia around the world [2,4,5]. For instance, as experimental activities, microstructural features and changes due to corrosion were examined [6,7], characteristics of outside diameter SCC (ODSCC) were assessed and mechanical properties were measured [8]. As numerical activities, mainly, effective finite element (FE) analysis methods were developed and used to predict weld residual stresses and expanded for fatigue life and fracture investigation [9,10].
Effects of Filler Metal Composition on Inclusions and Inclusion Defects for ER NiCrFe-7 Weldments
2013, Journal of Materials Science and TechnologyCitation Excerpt :Compared with Incoloy 800 and type 304 stainless steel, Inconel 690 shows good resistance against stress corrosion cracking in high temperature water, oxygenated and deoxygenated environment, and with or without crevices and chloride or lead contamination[1–4]. Thus, Inconel 690 has been employed in nuclear power plant components as a corrosion-resistant material instead of Inconel 600[5–8]. The matrix of Inconel 690 is austenite with semi-continuous to continuous carbides at grain boundaries and with a few intragranular carbides and nitrides[9–12].
Micro-segregation and Precipitation of Alloy 690 during Isothermal Solidification: The Role of Nitrogen Content
2012, Journal of Materials Science and TechnologyA study of life prediction differences for a nickel-base Alloy 690 using a threshold and a non-threshold model
2009, Journal of Nuclear MaterialsThe response of alloy 690 tubing in a pressurized water reactor environment
2007, Materials and DesignAn investigation of the fatigue crack growth behavior of INCONEL 690
2006, Materials Science and Engineering: A