Fatigue crack growth rates of irradiated pressure vessel steels in simulated nuclear coolant environment
Bibliographic record
Abstract
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Research context
OpenAlex topics: Fatigue and fracture mechanics · Nuclear Engineering Thermal-Hydraulics · Thermodynamic and Structural Properties of Metals and Alloys
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References (15)
Show source-provided references
1. Cullen · 1980
2. Section XI of the ASME Boiler and Pressure Vessel Code, Rules for In-service Inspection of Nuclear Power Plant Components ANST/ASME-BPV-XI-1 (American Society of Mechanical Engineers, New York, issued annually).
3. Clark · J. Mater. · 1971 · 16 · 134
4. Paris · ASTM-STP 513 · 1972 · 141
5. T.R. Mager and V.J. McLoughlin, HSST Technical Report 16, Westinghouse Report WCAP-7776.
6. Bamford · 1978
7. ASTM E647-78T, ASTM Book of Standards, Vol. 10 (issued annually).
8. 1968 · 29
9. Watson · 1976
10. Watson · 1977
11. Wei · 1971
12. Cullen · 1979
13. Cullen · 1979
14. Cullen · 1980
15. James · Nucl. Safety · 1973 · 18 · 791