Probability of failure in BWR (Boiling Water Reactor) reactor coolant piping: Volume 2, Pipe failure induced by crack growth and failure of intermediate supports

Technical Report · Wed Mar 01 04:00:00 EST 1989

DOI:https://doi.org/10.2172/6163101· OSTI ID:6163101

Lo, T.; Bumpus, S. E.; Chinn, D. J.; Mensing, R. W.; Holman, G. S.

The US Nuclear Regulatory Commission (NRC) contracted with the Lawrence Livermore National Laboratory (LLNL) to conduct a study to determine if the probability of occurrence of a double-ended guillotine break (DEGB) in the major coolant piping systems of nuclear power plants is large enough to warrant the current stringent design requirements of designing against the postulated effects of a DEGB. The study includes both the PWR (Pressurized Water Reactor) and the BWR (Boiling Water Reactor) plants in the United States. Following the study of PWR plants, a study of BWR reactor coolant piping was performed. The Brunswick Steam Electric Plant at Southport, North Carolina was selected as the pilot plant for the BWR evaluation. The probability of pipe failure in three major coolant pipings was assessed: the recirculation loops, the primary steam lines, and the main feedwater lines. In the case of recirculation loops, both the existing and a proposed replacement system were studied. A probabilistic fracture mechanics approach was used in this study to estimate the crack growth and to assess the crack stability in the piping systems throughout the lifetime of the plant. The effects of the failure of intermediate pipe supports were also examined. The results of the assessment indicated that the probability of occurrence of DEGB due to crack growth and instability is small if the problem of intergranular stress corrosion cracking (IGSCC) is resolved by the use of the replacement system. 33 refs., 41 figs., 32 tabs.

• Research Organization: Nuclear Regulatory Commission, Washington, DC (USA). Div. of Engineering; Lawrence Livermore National Lab., CA (USA)
• DOE Contract Number: W-7405-ENG-48
• OSTI ID: 6163101
• Report Number(s): NUREG/CR-4792-Vol.2; UCID-20914-Vol.2; ON: TI89008910
• Country of Publication: United States
• Language: English

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99 GENERAL AND MISCELLANEOUS
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ALLOYS
BWR TYPE REACTORS
COOLING SYSTEMS
CRACK PROPAGATION
CRACKS
ENERGY SYSTEMS
FAILURES
FATIGUE
FRACTURE MECHANICS
FRACTURE PROPERTIES
HIGH ALLOY STEELS
IRON ALLOYS
IRON BASE ALLOYS
JOINTS
LEAKS
MATHEMATICS
MECHANICAL PROPERTIES
MECHANICAL STRUCTURES
MECHANICS
PIPES
PROBABILITY
PWR TYPE REACTORS
REACTOR COMPONENTS
REACTOR COOLING SYSTEMS
REACTOR SAFETY
REACTORS
SAFETY
SEISMIC EFFECTS
STAINLESS STEELS
STATISTICS
STEELS
SUPPORTS
WATER COOLED REACTORS
WATER MODERATED REACTORS
WELDED JOINTS