Prediction of aging degradation of cast stainless steel components in LWR systems

Conference · Sun Mar 01 04:00:00 EST 1992

OSTI ID:10154056

Chopra, O K

A procedure and correlations are presented for predicting Charpy-impact energy, tensile flow stress, fracture toughness J-R curve, and J{sub IC} of aged cast stainless steels from known material information. The ``saturation`` impact strength and fracture toughness of a specific cast stainless steel, i.e., the minimum value that would be achieved for the material after long-term service, is estimated from the chemical composition of the steel. Mechanical properties as a function of time and temperature of reactor service are estimated from impact energy and flow stress of the unaged material and the kinetics of embrittlement, which are also determined from chemical composition. The J{sub IC} values are determined from the estimated J-R curve and flow stress. Examples of estimating mechanical properties of of cast stainless steel components during reactor service are presented. A common ``predicted lower-bound` J-R curve for cast stainless steels of unknown chemical composition is also defined for a given grade of steel, ferrite content, and temperature.

• Research Organization: Argonne National Lab., IL (United States)
• Sponsoring Organization: Nuclear Regulatory Commission, Washington, DC (United States)
• DOE Contract Number: W-31109-ENG-38
• OSTI ID: 10154056
• Report Number(s): ANL/CP--75791; CONF-920375--24; ON: DE92014848
• Country of Publication: United States
• Language: English

Related Subjects

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
210100
210200
36 MATERIALS SCIENCE
360103
360106
AGING
BWR TYPE REACTORS
EMBRITTLEMENT
FRACTURE PROPERTIES
MECHANICAL PROPERTIES
PHYSICAL RADIATION EFFECTS
POWER REACTORS
NONBREEDING
LIGHT-WATER MODERATED
BOILING WATER COOLED
POWER REACTORS
NONBREEDING
LIGHT-WATER MODERATED
NONBOILING WATER COOLED
PWR TYPE REACTORS
RADIATION EFFECTS
REACTOR COMPONENTS
STAINLESS STEELS