Assessment of short through-wall circumferential cracks in pipes. Experiments and analysis: March 1990--December 1994

Technical Report · Sat Apr 01 04:00:00 EST 1995

DOI:https://doi.org/10.2172/46619· OSTI ID:46619

Brust, F W; Scott, P; Rahman, S ^[1]

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1. Battelle, Columbus, OH (United States); and others

This topical report summarizes the work performed for the Nuclear Regulatory Commission`s (NRC) research program entitled ``Short Cracks in Piping and Piping Welds`` that specifically focuses on pipes with short through-wall cracks. Previous NRC efforts, conducted under the Degraded Piping Program, focused on understanding the fracture behavior of larger cracks in piping and fundamental fracture mechanics developments necessary for this technology. This report gives details on: (1) material property determinations, (2) pipe fracture experiments, and (3) development, modification, and validation of fracture analysis methods. The material property data required to analyze the experimental results are included. These data were also implemented into the NRC`s PIFRAC database. Three pipe experiments with short through-wall cracks were conducted on large diameter pipe. Also, experiments were conducted on a large-diameter uncracked pipe and a pipe with a moderate-size through-wall crack. The analysis results reported here focus on simple predictive methods based on the J-Tearing theory as well as limit-load and ASME Section 11 analyses. Some of these methods were improved for short-crack-length predictions. The accuracy of the various methods was determined by comparisons with experimental results from this and other programs. 69 refs., 124 figs, 49 tabs.

• Research Organization: Nuclear Regulatory Commission, Washington, DC (United States). Div. of Engineering Technology; Battelle, Columbus, OH (United States)
• Sponsoring Organization: Nuclear Regulatory Commission, Washington, DC (United States)
• DOE Contract Number: FG02-90ER14135
• OSTI ID: 46619
• Report Number(s): NUREG/CR--6235; BMI--2179; ON: TI95010506
• Country of Publication: United States
• Language: English

Related Subjects

22 GENERAL STUDIES OF NUCLEAR REACTORS
36 MATERIALS SCIENCE
CARBON STEELS
CHARPY TEST
CRACK PROPAGATION
EXPERIMENTAL DATA
FRACTURE PROPERTIES
LOSS OF COOLANT
NUCLEAR POWER PLANTS
PIPES
STAINLESS STEEL-316
STAINLESS STEEL-316L
TENSILE PROPERTIES
WELDED JOINTS