Extension and Extrapolation of J-R Curves and Their Application to the Low Upper Shelf Toughness Issue

Technical Report · Fri Mar 01 04:00:00 EST 1991

DOI:https://doi.org/10.2172/5876140· OSTI ID:5876140

Joyce, J. A. ^[1]; Hackett, E. M. ^[2]

---

1. US Naval Academy, Annapolis, MD (United States)
2. David Taylor Research Center, Annapolis, MD (United States)

This document develops methods of measuring experimentally the limits of valid fracture mechanics data that can be obtained from small fracture mechanics specimens. The proposed technique generally shows that present ASTM limits are overly conservative and the new technique would allow almost a three fold increase in the amount of crack extension allowed in the testing of a surveillance specimen. Analytic relationships are then developed to allow use of the new experimentally measured limit to J controlled crack growth for design or failure analysis applications to pressure vessel structures. The new region of J controlled crack growth is shown to correlate best with the omega criterion which defines limits on both the maximum J level and the maximum crack extension allowable for a particular specimen size and material toughness combination. The final section looks at the problem of extrapolation of J-R curve data when needed for a structure fracture analysis. Several forms of extrapolation relationships are compared from the point of view of accurate and conservative extrapolation, particularly from the standpoint of tearing instability analysis of a growing, ductile crack on the material upper shelf.

• Research Organization: US Nuclear Regulatory Commission (NRC), Washington, DC (United States). Div. of Engineering; US Naval Academy, Annapolis, MD (United States); David Taylor Research Center, Annapolis, MD (United States)
• Sponsoring Organization: USNRC
• OSTI ID: 5876140
• Report Number(s): NUREG/CR--5577; ON: TI91009862
• Country of Publication: United States
• Language: English

Related Subjects

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
210100* -- Power Reactors
Nonbreeding
Light-Water Moderated
Boiling Water Cooled
210200 -- Power Reactors
Nonbreeding
Light-Water Moderated
Nonboiling Water Cooled
ACCIDENTS
AGING
BWR TYPE REACTORS
CHARPY TEST
CONTAINERS
CRACK PROPAGATION
DESTRUCTIVE TESTING
ELASTICITY
FRACTURE MECHANICS
IMPACT TESTS
J-R curve
MATERIALS TESTING
MATHEMATICAL MODELS
MECHANICAL PROPERTIES
MECHANICAL TESTS
MECHANICS
PHYSICAL RADIATION EFFECTS
PLASTICITY
PRESSURE VESSELS
PWR TYPE REACTORS
RADIATION EFFECTS
REACTOR ACCIDENTS
REACTOR SAFETY
REACTORS
SAFETY
SERVICE LIFE
SURVEILLANCE
TENSILE PROPERTIES
TESTING
VALIDATION
WATER COOLED REACTORS
WATER MODERATED REACTORS
fracture mechanics