Stress Corrosion Cracking Behavior of Cast Stainless Steels
- University of Michigan
- ORNL
Casting of austenitic stainless steels offers the possibility of directly producing large and/or relatively complex structures, such as the first wall shield modules or the diverter cassette for the ITER fusion reactor. Casting offers major cost savings when compared to fabrication via welding of quarter modules machined from large forgings. However, the strength properties of such cast components are typically considered inferior to those of conventionally forged and annealed components. To improve and validate cast stainless steel as a substitute for wrought stainless steel, a development and testing program was initiated, utilizing nitrogen and manganese additions to promote improved performance. This paper focuses on the response of the first set of developmental alloys to neutron-irradiation and susceptibility to stress corrosion cracking. These cast materials may also have applications for different components in light water reactors. Results showed that all steels exhibited irradiation-induced hardening and a corresponding drop in ductility, as expected, although there is still considerable ductility in the irradiated samples. The cast steels all exhibited reduced hardening in comparison to a wrought reference steels, which may be related to a larger grain size. Higher nitrogen contents did not negatively influence irradiation performance. Regarding stress corrosion cracking susceptibility, the large difference in grain size limits the comparison between wrought and cast materials, and inclusions in a reference and archive cast alloy tests complicate analysis of these samples. Results suggest that the irradiated archive heat was more susceptible to cracking than the modified alloys, which may be related to the more complex microstructure. Further, the results suggest that the modified cast steel is at least as SCC resistant as wrought 316LN. The beneficial effect of nitrogen on the mechanical properties of the alloys remains after irradiation and is not detrimental to SCC resistance.
- Research Organization:
- Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)
- Sponsoring Organization:
- USDOE Office of Nuclear Energy (NE)
- DOE Contract Number:
- DE-AC05-00OR22725
- OSTI ID:
- 1014219
- Resource Relation:
- Conference: 14th International Conference on Environmen tal Degradation of Materials in Nuclear Power systems (2009), Virginia Beach, VA, VA, USA, 20090823, 20090827
- Country of Publication:
- United States
- Language:
- English
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Related Subjects
70 PLASMA PHYSICS AND FUSION TECHNOLOGY
ALLOYS
AUSTENITIC STEELS
CASTING
DUCTILITY
FABRICATION
FIRST WALL
GRAIN SIZE
HARDENING
IRRADIATION
MANGANESE ADDITIONS
MECHANICAL PROPERTIES
MICROSTRUCTURE
NITROGEN
NUCLEAR POWER
PERFORMANCE
SHIELDS
STAINLESS STEELS
STEELS
STRESS CORROSION
THERMONUCLEAR REACTORS