Long-term embrittlement of cast duplex stainless steels in LWR systems: Semiannual report, October 1986-March 1987
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 the laboratory or in reactor service have been characterized by various analytical techniques. The results indicate that the low-temperature embrittlement of the ferrite phase of the duplex structure is caused primarily by the formation of the Cr-rich ..cap alpha..' phase and the Ni- and Si-rich G phase. The ..cap alpha..' phase forms by spinodal decomposition; however, the nucleation and growth of the ..cap alpha..' phase is very slow in the temperature range of interest. The kinetics of G-phase precipitation depend on the molybdenum and carbon content of the steel. The extent of spinodal decomposition and G-phase precipitation is strongly influenced by temperature. The variation in precipitation behavior indicates that the data obtained from accelerated aging at greater than or equal to 400/sup 0/C cannot be extrapolated to reactor temperatures for predicting the end-of-life fracture toughness of reactor components. 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. Thermal aging of the cast materials results in an increase in yield and ultimate strengths and a decrease in impact energy, J/sub IC/, and tearing modulus. The relative reduction in J/sub IC/ values of the aged materials is similar to the relative decrease in impact energy. The carbon content of the steel is an important factor in the control of the overall process of embrittlement. The influence of compositional and metallurgical parameters on impact strength and fracture mode is discussed. 21 refs., 19 figs., 5 tabs.
- Research Organization:
- Argonne National Lab., IL (USA)
- DOE Contract Number:
- W-31109-ENG-38
- OSTI ID:
- 5808201
- Report Number(s):
- NUREG/CR-4744-Vol.2-No.1; ANL-87-45; ON: TI88003538
- Country of Publication:
- United States
- Language:
- English
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Related Subjects
36 MATERIALS SCIENCE
BWR TYPE REACTORS
REACTOR COMPONENTS
PWR TYPE REACTORS
STAINLESS STEELS
AGING
EMBRITTLEMENT
THERMAL DEGRADATION
ANNEALING
CARBON
CHROMIUM
FERRITE
MICROSTRUCTURE
MOLYBDENUM
PROGRESS REPORT
RESEARCH PROGRAMS
ALLOYS
CARBON ADDITIONS
CHROMIUM ALLOYS
CORROSION RESISTANT ALLOYS
CRYSTAL STRUCTURE
DOCUMENT TYPES
ELEMENTS
HEAT TREATMENTS
IRON ALLOYS
IRON BASE ALLOYS
METALS
NONMETALS
REACTORS
STEELS
TRANSITION ELEMENTS
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
360103 - Metals & Alloys- Mechanical Properties