Fatigue crack propagation of low-alloy steel in a vacuum environment: Final report
Technical Report
·
OSTI ID:6119103
The fatigue-crack propagation (FCP) behavior of ASTM A533-B-1 steel was characterized in vacuo at 288/sup 0/C. Tests were conducted at two stress ratios: R = 0.05 and R = 0.7. Results of these tests were compared with results from precious studies for the same type of steel tested in an air environment, and FCP rates in vacuo were generally lower than those in air. Stress ratio effects in vacuo were not as great as those in air, and both stress ratio effects and environmental effects are discussed from the standpoint of crack closure concepts.
- Research Organization:
- Westinghouse Hanford Co., Richland, WA (USA); Electric Power Research Inst., Palo Alto, CA (USA)
- DOE Contract Number:
- AC06-76FF02170
- OSTI ID:
- 6119103
- Report Number(s):
- EPRI-NP-5345; ON: DE87013770
- Country of Publication:
- United States
- Language:
- English
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Related Subjects
22 GENERAL STUDIES OF NUCLEAR REACTORS
220200 -- Nuclear Reactor Technology-- Components & Accessories
36 MATERIALS SCIENCE
360103* -- Metals & Alloys-- Mechanical Properties
ALLOYS
CARBON STEELS
CONTAINERS
CRACK PROPAGATION
FATIGUE
HIGH TEMPERATURE
HIGH VACUUM
IRON ALLOYS
IRON BASE ALLOYS
MATERIALS TESTING
MECHANICAL PROPERTIES
NUCLEAR FACILITIES
NUCLEAR POWER PLANTS
POWER PLANTS
PRESSURE VESSELS
STEEL-ASTM-A533-B
STEELS
TESTING
THERMAL POWER PLANTS
220200 -- Nuclear Reactor Technology-- Components & Accessories
36 MATERIALS SCIENCE
360103* -- Metals & Alloys-- Mechanical Properties
ALLOYS
CARBON STEELS
CONTAINERS
CRACK PROPAGATION
FATIGUE
HIGH TEMPERATURE
HIGH VACUUM
IRON ALLOYS
IRON BASE ALLOYS
MATERIALS TESTING
MECHANICAL PROPERTIES
NUCLEAR FACILITIES
NUCLEAR POWER PLANTS
POWER PLANTS
PRESSURE VESSELS
STEEL-ASTM-A533-B
STEELS
TESTING
THERMAL POWER PLANTS