scholarly journals Strength and ductility of fast reactor irradiated austenitic stainless steels

1979 ◽  
Author(s):  
L D Blackburn
Keyword(s):  
Fast Reactor ◽  
Stainless Steels ◽  
Austenitic Stainless Steels ◽  
Strength And Ductility

scholarly journals Effect of Hydrogen on Tensile Strength and Ductility of Multi-Pass 304L / 308L Austenitic Stainless Steel Welds

2015 ◽  
Author(s):  
Dorian K. Balch ◽  
Chris San Marchi
Keyword(s):  
Tensile Strength ◽  
Stainless Steels ◽  
Tensile Elongation ◽  
Austenitic Stainless Steels ◽  
Tensile Specimen ◽  
Rolled Plate ◽  
Gas Tungsten Arc ◽  
Steel Welds ◽  
Weld Root ◽  
Strength And Ductility

Austenitic stainless steels such as 304L are frequently used for hydrogen service applications due to their excellent resistance to hydrogen embrittlement. However, welds in austenitic stainless steels often contain microstructures that are more susceptible to the presence of hydrogen. This study examines the tensile strength and ductility of a multi-pass gas tungsten arc weld made on 304L cross-rolled plate using 308L weld filler wire. Sub-sized tensile specimens were used to ensure the entire gage section of each tensile specimen consisted of weld metal. Specimens were extracted in both axial and transverse orientations, and at three different depths within the weld (root, center, and top). Yield strength decreased and ductility increased moving from the root to the top of the weld. A subset of specimens was precharged with hydrogen at 138 MPa (20,000 psi) and 300°C prior to testing, resulting in a uniform hydrogen concentration of 7700 appm. The presence of hydrogen resulted in a slight increase in yield and tensile strength and a roughly 50% decrease in tensile elongation and reduction in area, compared to the hydrogen-free properties.

Austenitic stainless steels of high strength and ductility

2004 ◽  
Vol 95 (7) ◽  
pp. 596-600 ◽  
Author(s):  
Markus O. Speidel ◽  
Hannes J. C. Speidel
Keyword(s):  
Stainless Steels ◽  
Austenitic Stainless Steels ◽  
High Strength ◽  
Strength And Ductility

Microstructural Evolution of Austenitic Stainless Steels Irradiated in a Fast Reactor

2009 ◽  
pp. 817-817-14 ◽  
Author(s):  
OV Borodin ◽  
VV Bryk ◽  
VN Voyevodin ◽  
IM Neklyudov ◽  
VK Shamardin ◽  
...  
Keyword(s):  
Microstructural Evolution ◽  
Fast Reactor ◽  
Stainless Steels ◽  
Austenitic Stainless Steels

A Comparative Study of Negligible Creep Curves for Rational Elevated Temperature Design

2009 ◽  
Author(s):  
Masanori Ando ◽  
Nobuhiro Isobe ◽  
Nobuchika Kawasaki ◽  
Masayuki Sukekawa ◽  
Naoto Kasahara
Keyword(s):  
Elevated Temperature ◽  
Creep Curve ◽  
Fast Reactor ◽  
Stainless Steels ◽  
Material Properties ◽  
Austenitic Stainless Steels ◽  
The Other ◽  
Evaluation Procedure ◽  
Creep Curves ◽  
Creep Time

In Japanese elevated temperature standard, creep considering design is required for all ferrite steels applied over 375°C and all austenitic stainless steels applied over 425°C regardless of the operating time. On the other hands, ASME Sec.III Subsection NH, RCC-MR and R5 provide the additional rules to determine the negligible creep range. In those standards, each material is evaluated as non-creep considering design region, although there are varieties of applicable materials and the rules to settle the negligible creep range in each standard. 316FR and Mod.9Cr-1Mo are candidate materials of Japan sodium-cooled fast reactor (JSFR), and those high creep resistant properties extend the negligible creep damage area over the conventional temperature limits. Extension of non-creep design area widens design windows and simplifies the creep analysis procedure. To reply those requirements, authors already proposed original negligible criterion and discussed about it. In this paper, we recall the backgrounds of the negligible creep criterion which have already been proposed. Then the negligible creep criterion and relating property in each standard were compared. For estimating the evaluation procedure of each criterion, the common material properties used in “Elevated Temperature Structural Design Guide for Commercialized Fast Reactor (FDS)” were applied to each standard’s criteria. All standards have the negligible creep curves/regions for type 18Cr-8Ni steels and type 18Cr-12Ni-2.5Mo steels, although ASME Sec.III Subsection NH defines just the criteria of negligible creep for the rule of inelastic strain limits. On the diagram of temperature-negligible creep time, the negligible creep curves of 316L(N)(1S) in RCC-MR and R5 exist between those of SUS316 and 316FR in FDS. The negligible creep regions defined in all standards are similar for austenitic stainless steels, although those criteria are different. Comparison of the negligible creep curves by each criterion with FDS’s material properties indicated that the criterion in FDS provides the most conservative curve. In case of Mod.9Cr-1Mo steel, FDS and R5 provide relationship between temperatures and time for estimating the negligible creep time. ASME Sec. III Subsection NH provides only procedures and has no practical allowable values, and RCC-MR doesn’t have the negligible creep curve. Comparison of the negligible creep curves in each criterion with FDS’s material properties indicated that FDS’s criterion allows the longest negligible creep. The negligible creep criteria in ASME Sec.III Subsection NH, RCC-MR and R5 are not practicable for Mod.9Cr-1Mo. On the other hands, FDS criterion raises the temperature limits from conventional 375°C to about 425°C even when the components designed lifetime is 60years. Sensitivities to the difference of criteria and material properties were discussed and concluded that negligible creep curve is strongly dependent on the combination of criteria and material properties. Some evaluations proved that the negligible creep curves in FDS are moderately conservative and practicable.

Radiation Damage Studies with the HVEM

1972 ◽  
Vol 30 ◽  
pp. 680-681
Author(s):  
J. J. Laidler ◽  
B. Mastel
Keyword(s):  
Radiation Damage ◽  
Stainless Steels ◽  
Volume Expansion ◽  
Austenitic Stainless Steels ◽  
Test Facility ◽  
Fast Breeder Reactors ◽  
Breeder Reactors ◽  
Under Construction ◽  
Development Laboratory ◽  
Insight Into

One of the major materials problems encountered in the development of fast breeder reactors for commercial power generation is the phenomenon of swelling in core structural components and fuel cladding. This volume expansion, which is due to the retention of lattice vacancies by agglomeration into large polyhedral clusters (voids), may amount to ten percent or greater at goal fluences in some austenitic stainless steels. From a design standpoint, this is an undesirable situation, and it is necessary to obtain experimental confirmation that such excessive volume expansion will not occur in materials selected for core applications in the Fast Flux Test Facility, the prototypic LMFBR now under construction at the Hanford Engineering Development Laboratory (HEDL). The HEDL JEM-1000 1 MeV electron microscope is being used to provide an insight into trends of radiation damage accumulation in stainless steels, since it is possible to produce atom displacements at an accelerated rate with 1 MeV electrons, while the specimen is under continuous observation.

Sign in / Sign up

Export Citation Format

Share Document