Calculating the tearing resistance of ductile steels: Final report
Technical Report
·
OSTI ID:6238978
This work was motivated by the results of fracture mechanics analysis for the low upper shelf (A-11) reactor vessel issue. The analysis indicated a need for J-resistance curve that includes crack extensions from about 0.5 to 1.0 inch. Specimens from reactor vessel surveillance capsules will not provide the needed crack extension. Therefore, there is a need for a reliable method that can be used to extrapolate small specimen J-resistance curve for evaluation of large components. This report focuses on the development of accurate J solution, assesses the impact of approximations on the current method, and proposes a criterion for extrapolating the J-resistance curve. This report presents new J-integral solutions for the compact tension (C(T)) and three-point bend bar (SE(B)) specimens. The solutions cover a wide range of crack lengths, allowing analysis for crack lengths greater than 20 percent of the specimen width. These solutions are useful for generating accurate J-resistance curves. Solutions for both the J/sub d/ and J/sub M/ are derived. The digitized form of the J/sub d/ and J/sub M/ solutions suitable for implementing into a computer code are presented in an Appendix. 20 refs., 19 figs., 7 tabs.
- Research Organization:
- Electric Power Research Inst., Palo Alto, CA (USA); NOVETECH Corp., Rockville, MD (USA)
- OSTI ID:
- 6238978
- Report Number(s):
- EPRI-NP-6310
- Country of Publication:
- United States
- Language:
- English
Similar Records
Extension and Extrapolation of J-R Curves and Their Application to the Low Upper Shelf Toughness Issue
Predictions of J-R curves with large crack growth from small specimen data
J-R curve characterization of irradiated low-shelf nuclear vessel steels
Technical Report
·
Thu Feb 28 23:00:00 EST 1991
·
OSTI ID:5876140
Predictions of J-R curves with large crack growth from small specimen data
Technical Report
·
Mon Sep 01 00:00:00 EDT 1986
·
OSTI ID:5049984
J-R curve characterization of irradiated low-shelf nuclear vessel steels
Conference
·
Thu Oct 01 00:00:00 EDT 1981
· Am. Soc. Test. Mater., Spec. Tech. Publ.; (United States)
·
OSTI ID:6410156
Related Subjects
22 GENERAL STUDIES OF NUCLEAR REACTORS
220200 -- Nuclear Reactor Technology-- Components & Accessories
36 MATERIALS SCIENCE
360103* -- Metals & Alloys-- Mechanical Properties
99 GENERAL AND MISCELLANEOUS
990230 -- Mathematics & Mathematical Models-- (1987-1989)
ALLOYS
AUSTENITIC STEELS
CHROMIUM ALLOYS
CHROMIUM-NICKEL STEELS
CONTAINERS
CORROSION RESISTANT ALLOYS
CRACK PROPAGATION
DOCUMENT TYPES
FRACTURE MECHANICS
HEAT RESISTANT MATERIALS
HEAT RESISTING ALLOYS
HIGH ALLOY STEELS
IRON ALLOYS
IRON BASE ALLOYS
MATERIALS
MATHEMATICAL MODELS
MECHANICS
NICKEL ALLOYS
PIPES
PRESSURE VESSELS
PROGRESS REPORT
RECOMMENDATIONS
STAINLESS STEEL-304
STAINLESS STEELS
STEEL-CR19NI10
STEELS
220200 -- Nuclear Reactor Technology-- Components & Accessories
36 MATERIALS SCIENCE
360103* -- Metals & Alloys-- Mechanical Properties
99 GENERAL AND MISCELLANEOUS
990230 -- Mathematics & Mathematical Models-- (1987-1989)
ALLOYS
AUSTENITIC STEELS
CHROMIUM ALLOYS
CHROMIUM-NICKEL STEELS
CONTAINERS
CORROSION RESISTANT ALLOYS
CRACK PROPAGATION
DOCUMENT TYPES
FRACTURE MECHANICS
HEAT RESISTANT MATERIALS
HEAT RESISTING ALLOYS
HIGH ALLOY STEELS
IRON ALLOYS
IRON BASE ALLOYS
MATERIALS
MATHEMATICAL MODELS
MECHANICS
NICKEL ALLOYS
PIPES
PRESSURE VESSELS
PROGRESS REPORT
RECOMMENDATIONS
STAINLESS STEEL-304
STAINLESS STEELS
STEEL-CR19NI10
STEELS