Interim fatigue design curves for carbon, low-alloy, and austenitic stainless steels in LWR environments
Conference
·
OSTI ID:6895817
Both temperature and oxygen affect fatigue life; at the very low dissolved-oxygen levels in PWRs and BWRs with hydrogen water chemistry, environmental effects on fatigue life are modest at all temperatures (T) and strain rates. Between 0.1 and 0.2 ppM, the effect of dissolved-oxygen increases rapidly. In oxygenated environments, fatigue life depends strongly on strain rate and T. A fracture mechanics model is developed for predicting fatigue lives, and interim environmentally assisted cracking (EAC)-adjusted fatigue curves are proposed for carbon steels, low-alloy steels, and austenitic stainless steels.
- Research Organization:
- Argonne National Lab., IL (United States)
- Sponsoring Organization:
- USNRC; Nuclear Regulatory Commission, Washington, DC (United States)
- DOE Contract Number:
- W-31109-ENG-38
- OSTI ID:
- 6895817
- Report Number(s):
- ANL/MCT/CP-78588; CONF-921007-23; ON: DE93009202; TRN: 93-009761
- Resource Relation:
- Conference: 20. water reactor safety information meeting, Bethesda, MD (United States), 21-23 Oct 1992
- Country of Publication:
- United States
- Language:
- English
Similar Records
Interim fatigue design curves for carbon, low-alloy, and austenitic stainless steels in LWR environments
Interim fatigue design curves for carbon, low-alloy, and austenitic stainless steels in LWR environments
Interim fatigue design curves for carbon, low-alloy, and austenitic stainless steels in LWR environments
Conference
·
Fri Jan 01 00:00:00 EST 1993
·
OSTI ID:6895817
Interim fatigue design curves for carbon, low-alloy, and austenitic stainless steels in LWR environments
Technical Report
·
Thu Apr 01 00:00:00 EST 1993
·
OSTI ID:6895817
Interim fatigue design curves for carbon, low-alloy, and austenitic stainless steels in LWR environments
Technical Report
·
Thu Apr 01 00:00:00 EST 1993
·
OSTI ID:6895817
Related Subjects
36 MATERIALS SCIENCE
22 GENERAL STUDIES OF NUCLEAR REACTORS
AUSTENITIC STEELS
FATIGUE
CARBON STEELS
LOW ALLOY STEELS
STAINLESS STEELS
BWR TYPE REACTORS
CORRELATIONS
CRACK PROPAGATION
FRACTURE MECHANICS
OXYGEN
PWR TYPE REACTORS
RECOMMENDATIONS
STRAIN RATE
TEMPERATURE RANGE 0400-1000 K
ALLOYS
ELEMENTS
ENRICHED URANIUM REACTORS
HIGH ALLOY STEELS
IRON ALLOYS
IRON BASE ALLOYS
MECHANICAL PROPERTIES
MECHANICS
NONMETALS
POWER REACTORS
REACTORS
STEELS
TEMPERATURE RANGE
THERMAL REACTORS
WATER COOLED REACTORS
WATER MODERATED REACTORS
360103* - Metals & Alloys- Mechanical Properties
220200 - Nuclear Reactor Technology- Components & Accessories
22 GENERAL STUDIES OF NUCLEAR REACTORS
AUSTENITIC STEELS
FATIGUE
CARBON STEELS
LOW ALLOY STEELS
STAINLESS STEELS
BWR TYPE REACTORS
CORRELATIONS
CRACK PROPAGATION
FRACTURE MECHANICS
OXYGEN
PWR TYPE REACTORS
RECOMMENDATIONS
STRAIN RATE
TEMPERATURE RANGE 0400-1000 K
ALLOYS
ELEMENTS
ENRICHED URANIUM REACTORS
HIGH ALLOY STEELS
IRON ALLOYS
IRON BASE ALLOYS
MECHANICAL PROPERTIES
MECHANICS
NONMETALS
POWER REACTORS
REACTORS
STEELS
TEMPERATURE RANGE
THERMAL REACTORS
WATER COOLED REACTORS
WATER MODERATED REACTORS
360103* - Metals & Alloys- Mechanical Properties
220200 - Nuclear Reactor Technology- Components & Accessories