Aging degradation of cast stainless steel
A program is being conducted to investigate the significance of in-service embrittlement of cast duplex stainless steels under light-water reactor operating conditions. Microstructures of cast materials subjected to long-term aging either in reactor service or in the laboratory have been characterized by TEM, SANS, and APFIM techniques. Two precipitate phases, i.e., the Cr-rich ..cap alpha..' and Ni- and Si-rich G phase, have been identified in the ferrite matrix of the aged steels. The results indicate that the low-temperature embrittlement is primarily caused by ..cap alpha..' precipitates which form by spinodal decomposition. The relative contribution of G phase to loss of toughness is now known. Microstructural data also indicate that weakening of ferrite/austenite phase boundary by carbide precipitates has a significant effect on the onset and extent of embrittlement of the high-carbon CF-8 and CF-8M grades of stainless steels, particularly after aging at 400 or 450/sup 0/C. Data from Charpy-impact, tensile, and J-R curve tests for several heats of cast stainless steel aged up to 10,000 h at 350, 400, and 450/sup 0/C are presented and correlated with the microstructural results. Thermal aging of the steels results in an increase in tensile strength and a decrease in impact energy, J/sub IC/, and tearing modulus. The fracture toughness results show good agreement with the Charpy-impact data. The effects of compositional and metallurgical variables on loss of toughness are discussed.
- Research Organization:
- Argonne National Lab., IL (USA)
- DOE Contract Number:
- W-31109-ENG-38
- OSTI ID:
- 7181088
- Report Number(s):
- CONF-8610135-59; ON: DE87004939
- Resource Relation:
- Conference: 14. water reactor safety information meeting, Gaithersburg, MD, USA, 27 Oct 1986; Other Information: Portions of this document are illegible in microfiche products
- Country of Publication:
- United States
- Language:
- English
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Related Subjects
36 MATERIALS SCIENCE
BWR TYPE REACTORS
REACTOR MATERIALS
PWR TYPE REACTORS
AGING
EMBRITTLEMENT
MATERIALS TESTING
MECHANICAL PROPERTIES
MICROSTRUCTURE
PHYSICAL RADIATION EFFECTS
RESEARCH PROGRAMS
STAINLESS STEELS
ALLOYS
CHROMIUM ALLOYS
CORROSION RESISTANT ALLOYS
CRYSTAL STRUCTURE
IRON ALLOYS
IRON BASE ALLOYS
MATERIALS
RADIATION EFFECTS
REACTORS
STEELS
TESTING
WATER COOLED REACTORS
WATER MODERATED REACTORS
210100* - Power Reactors
Nonbreeding
Light-Water Moderated
Boiling Water Cooled
210200 - Power Reactors
Nonbreeding
Light-Water Moderated
Nonboiling Water Cooled
360106 - Metals & Alloys- Radiation Effects
360103 - Metals & Alloys- Mechanical Properties