An investigation into the room temperature mechanical properties of nanocrystalline austenitic stainless steels

2013 ◽  
Vol 45 ◽  
pp. 674-681 ◽  
Author(s):  
Mostafa Eskandari ◽  
Abbas Zarei-Hanzaki ◽  
Hamid Reza Abedi
Keyword(s):  
Mechanical Properties ◽  
Stainless Steels ◽  
Room Temperature ◽  
Austenitic Stainless Steels

ALZ 316

Alloy Digest ◽  
1999 ◽  
Vol 48 (8) ◽  
Keyword(s):  
Mechanical Properties ◽  
Stainless Steel ◽  
Corrosion Resistance ◽  
Austenitic Stainless Steel ◽  
Physical Properties ◽  
Tensile Properties ◽  
Stainless Steels ◽  
Austenitic Stainless Steels ◽  
Heat Treating

Abstract ALZ 316 is an austenitic stainless steel with good formability, corrosion resistance, toughness, and mechanical properties. It is the basic grade of the stainless steels, containing 2 to 3% molybdenum. After the 304 series, the molybdenum-containing stainless steels are the most widely used austenitic stainless steels. This datasheet provides information on composition, physical properties, hardness, elasticity, and tensile properties. It also includes information on corrosion resistance as well as forming, heat treating, and joining. Filing Code: SS-756. Producer or source: ALZ nv.

scholarly journals Effect of Neutron Irradiation on the Mechanical Properties, Swelling and Creep of Austenitic Stainless Steels

Materials ◽  
2021 ◽  
Vol 14 (10) ◽  
pp. 2622
Author(s):  
Malcolm Griffiths
Keyword(s):  
Mechanical Properties ◽  
Stainless Steels ◽  
Dimensional Stability ◽  
Operating Experience ◽  
Austenitic Stainless Steels ◽  
High Strength ◽  
Gas Production ◽  
Rate Theory ◽  
Fast Reactors ◽  
Internal Structures

Austenitic stainless steels are used for core internal structures in sodium-cooled fast reactors (SFRs) and light-water reactors (LWRs) because of their high strength and retained toughness after irradiation (up to 80 dpa in LWRs), unlike ferritic steels that are embrittled at low doses (<1 dpa). For fast reactors, operating temperatures vary from 400 to 550 °C for the internal structures and up to 650 °C for the fuel cladding. The internal structures of the LWRs operate at temperatures between approximately 270 and 320 °C although some parts can be hotter (more than 400 °C) because of localised nuclear heating. The ongoing operability relies on being able to understand and predict how the mechanical properties and dimensional stability change over extended periods of operation. Test reactor irradiations and power reactor operating experience over more than 50 years has resulted in the accumulation of a large amount of data from which one can assess the effects of irradiation on the properties of austenitic stainless steels. The effect of irradiation on the intrinsic mechanical properties (strength, ductility, toughness, etc.) and dimensional stability derived from in- and out-reactor (post-irradiation) measurements and tests will be described and discussed. The main observations will be assessed using radiation damage and gas production models. Rate theory models will be used to show how the microstructural changes during irradiation affect mechanical properties and dimensional stability.

scholarly journals EVALUATION OF THE METALLURGICAL PARAMETERS EFFECT ON TENSILE PROPERTIES IN AUSTENITIC STAINLESS STEELS

2017 ◽  
Vol 23 (2) ◽  
pp. 111 ◽  
Author(s):  
Andrea Di Schino ◽  
Maria Richetta
Keyword(s):  
Mechanical Properties ◽  
Tensile Properties ◽  
Stainless Steels ◽  
Austenitic Stainless Steels ◽  
304 Stainless Steel ◽  
Strengthening Effect ◽  
Regression Equations ◽  
Aisi 304 ◽  
Solid Solution Strengthening ◽  
Cold Rolling And Annealing

<p>Even if relations predicting the mechanical properties on bars of austenitic stainless steels are already available, but no systematic works was carried out in order to predict mechanical properties in after cold rolling and annealing.   The tensile properties of a large number of cold rolled and annealed AISI 304 stainless steel are here correlated with their chemical composition and microstructure. Quantitative effects of various strengthening mechanisms such as grain size, d– ferrite content and solid solution strengthening by both interstitial and substitutional solutes are described. Interstitial solutes have by far the greatest strengthening effect and, among the substitutional solutes, the ferrite – stabilising elements have a greater effect than the austenite – stabilising elements. Regression equations are developed which predict with good accuracy the proof stress and tensile strength in AISI 304 stainless steels.</p>

Derivation of Design Fatigue Curves for Austenitic Stainless Steel Grades 1.4541 and 1.4550 Within the German Nuclear Safety Standard KTA 3201.2

2013 ◽  
Author(s):  
Xaver Schuler ◽  
Karl-Heinz Herter ◽  
Jürgen Rudolph
Keyword(s):  
Austenitic Stainless Steel ◽  
Stainless Steels ◽  
Room Temperature ◽  
Elevated Temperatures ◽  
Austenitic Stainless Steels ◽  
Nuclear Safety ◽  
Safety Standard ◽  
Fatigue Design ◽  
Steel Grades ◽  
Best Fit

Titanium and niobium stabilized austenitic stainless steels X6CrNiTi18-10S (material number 1.4541, correspondent to Alloy 321) respectively X6CrNiNb18-10S (material number 1.4550, correspondent to Alloy 347) are widely applied materials in German nuclear power plant components. Related requirements are defined in Nuclear Safety Standard KTA 3201.1. Fatigue design analysis is based on Nuclear Safety Standard KTA 3201.2. The fatigue design curve for austenitic stainless steels in the current valid edition of KTA 3201.2 is essentially identical with the design curve included in ASME-BPVC III, App I (ed. 2007, add. July 2008 respectively back editions). In the current code revision activities of KTA 3201.2 the compatibility of latest in air fatigue data for austenitic stainless steels with the above mentioned grades were examined in detail. The examinations were based on statistical evaluations of 149 strain controlled test data at room temperature and 129 data at elevated temperatures to derive best-fit mean data curves. Results of two additional load controlled test series (at room temperature and 288°C) in the high cycle regime were used to determine a technical endurance limit at 107 cycles. The related strain amplitudes were determined by consideration of the cyclic stress strain curve. The available fatigue data for the two austenitic materials at room temperature and elevated temperatures showed a clear temperature dependence in the high cycle regime demanding for two different best-fit curves. The correlation of the technical endurance limit(s) at room temperature and elevated temperatures with the ultimate strength of the materials is discussed. Design fatigue curves were derived by application of the well known factors to the best-fit curves. A factor of SN = 12 was applied to load cycles correspondent to the NUREG/CR-6909 approach covering influences of data scatter, surface roughness, size and sequence. In terms of strain respectively stress amplitudes in the high cycle regime, for elevated temperatures (>80°C) a factor of Sσ = 1.79 was applied considering and combining in detail the partial influences of data scatter surface roughness, size and mean stress. For room temperature a factor of Sσ = 1.88 shall be applied. As a result, new design fatigue curves for austenitic stainless steel grades 1.4541 and 1.4550 will be available within the German Nuclear Safety Standard KTA 3201.2. The fatigue design rules for all other austenitic stainless steel grades will be based on the new ASME-BPVC III, App I (ed. 2010) design curve.

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