Short cracks in piping and piping welds. Seventh program report, March 1993-December 1994. Volume 4, Number 1

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

DOI:https://doi.org/10.2172/50936· OSTI ID:50936

Wilkowski, G M; Ghadiali, N; Rudland, D; Krishnaswamy, P; Rahman, S; Scott, P ^[1]

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1. Battelle, Columbus, OH (United States)

This is the seventh progress report of the U.S. Nuclear Regulatory Commission`s research program entitled {open_quotes}Short Cracks in Piping and Piping Welds{close_quotes}. The program objective is to verify and improve fracture analyses for circumferentially cracked large-diameter nuclear piping with crack sizes typically used in leak-before-break (LBB) analyses and in-service flaw evaluations. All work in the eight technical tasks have been completed. Ten topical reports are scheduled to be published. Progress only during the reporting period, March 1993 - December 1994, not covered in the topical reports is presented in this report. Details about the following efforts are covered in this report: (1) Improvements to the two computer programs NRCPIPE and NRCPIPES to assess the failure behavior of circumferential through-wall and surface-cracked pipe, respectively; (2) Pipe material property database PIFRAC; (3) Circumferentially cracked pipe database CIRCUMCK.WKI; (4) An assessment of the proposed ASME Section III design stress rule changes on pipe flaw tolerance; and (5) A pipe fracture experiment on a section of pipe removed from service degraded by microbiologically induced corrosion (MIC) which contained a girth weld crack. Progress in the other tasks is not repeated here as it has been covered in great detail in the topical reports.

• Research Organization: Nuclear Regulatory Commission, Washington, DC (United States). Div. of Engineering Technology
• Sponsoring Organization: Nuclear Regulatory Commission, Washington, DC (United States)
• OSTI ID: 50936
• Report Number(s): NUREG/CR--4599-Vol.4-No.1; BMI--2173-Vol.4-No.1; ON: TI95010955
• Country of Publication: United States
• Language: English

Related Subjects

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
22 GENERAL STUDIES OF NUCLEAR REACTORS
36 MATERIALS SCIENCE
C CODES
CRACKS
FRACTURE PROPERTIES
HYDRAULICS
N CODES
P CODES
PIPES
PRIMARY COOLANT CIRCUITS
STAINLESS STEEL-304
STEEL-ASTM-A106
STRESS ANALYSIS
US NRC
WATER COOLED REACTORS
WELDED JOINTS