Refinement and evaluation of crack-opening-area analyses for circumferential through-wall cracks in pipes

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

DOI:https://doi.org/10.2172/46620· OSTI ID:46620

Rahman, S; Brust, F; Ghadiali, N; Krishnaswamy, P; Wilkowski, G ^[1]; Choi, Y H ^[1]; Moberg, F; Brickstad, B ^[1]

---

1. Battelle, Columbus, OH (United States)

Leak-before-break (LBB) analyses for circumferentially cracked pipes are currently being conducted in the nuclear industry to justify elimination of pipe whip restraints and jet impingement shields which are present because of the expected dynamic effects from pipe rupture. The application of the LBB methodology frequently requires calculation of leak rates. These leak rates depend on the crack-opening area of a through-wall crack in the pipe. In addition to LBB analyses, which assume a hypothetical flaw size, there is also interest in the integrity of actual leaking cracks corresponding to current leakage detection requirements in NRC Regulatory Guide 1.45, or for assessing temporary repair of Class 2 and 3 pipes that have leaks as are being evaluated in ASME Section 11. This study was requested by the NRC to review, evaluate, and refine current analytical models for crack-opening-area analyses of pipes with circumferential through-wall cracks. Twenty-five pipe experiments were analyzed to determine the accuracy of the predictive models. Several practical aspects of crack-opening such as; crack-face pressure, off-center cracks, restraint of pressure-induced bending, cracks in thickness transition regions, weld residual stresses, crack-morphology models, and thermal-hydraulic analysis, were also investigated. 140 refs., 105 figs., 41 tabs.

• Research Organization: US Nuclear Regulatory Commission (NRC), Washington, DC (United States). Div. of Engineering Technology; Battelle Memorial Inst., Columbus, OH (United States)
• Sponsoring Organization: Nuclear Regulatory Commission, Washington, DC (United States)
• DOE Contract Number: FG02-90ER14135
• OSTI ID: 46620
• Report Number(s): NUREG/CR-6300; BMI-2184; ON: TI95010503; TRN: 95:011068
• Resource Relation: Other Information: PBD: Apr 1995
• Country of Publication: United States
• Language: English

Related Subjects

22 NUCLEAR REACTOR TECHNOLOGY
36 MATERIALS SCIENCE
NUCLEAR POWER PLANTS
PIPES
CRACK PROPAGATION
MATHEMATICAL MODELS
LEAKS
LOSS OF COOLANT
CRACKS
MORPHOLOGY
PRESSURE DEPENDENCE
THERMAL ANALYSIS
HYDRAULICS
FRACTURE MECHANICS
ELASTICITY
PLASTICITY
COMPARATIVE EVALUATIONS
SENSITIVITY ANALYSIS
RESTRAINTS
COMPUTER CODES
EXPERIMENTAL DATA