Probability of pipe fracture in the primary coolant loop of a PWR plant. Volume 5. Probabilistic fracture mechanics analysis. Load Combination Program Project I final report
The primary purpose of the Load Combination Program covered in this report is to estimate the probability of a seismic induced LOCA in the primary piping of a commercial pressurized water reactor (PWR). Best estimates, rather than upper bound results are desired. This was accomplished by use of a fracture mechanics model that employs a random distribution of initial cracks in the piping welds. Estimates of the probability of cracks of various sizes initially existing in the welds are combined with fracture mechanics calculations of how these cracks would grow during service. This then leads to direct estimates of the probability of failure as a function of time and location within the piping system. The influence of varying the stress history to which the piping is subjected is easily determined. Seismic events enter into the analysis through the stresses they impose on the pipes. Hence, the influence of various seismic events on the piping failure probability can be determined, thereby providing the desired information.
- Research Organization:
- Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Science Applications International Corp., La Jolla, CA (USA)
- DOE Contract Number:
- W-7405-ENG-48
- OSTI ID:
- 5341468
- Report Number(s):
- NUREG/CR-2189-Vol.5; UCID-18967-Vol.5; ON: TI85016245
- Country of Publication:
- United States
- Language:
- English
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Probability of pipe fracture in the primary coolant loop of a PWR plant. Volume 9. PRAISE computer code user's manual. Load Combination Program Project I final report
Probability of pipe fracture in the primary coolant loop of a PWR plant. Volume 5: probabilistic fracture mechanics analysis. Final report
Related Subjects
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
LOSS OF COOLANT
DYNAMIC LOADS
PIPES
FAILURES
FRACTURE MECHANICS
PRIMARY COOLANT CIRCUITS
PWR TYPE REACTORS
SEISMIC EFFECTS
CRACKS
REACTOR SAFETY
STRESS ANALYSIS
WELDED JOINTS
ACCIDENTS
COOLING SYSTEMS
ENERGY SYSTEMS
JOINTS
MECHANICS
REACTOR ACCIDENTS
REACTOR COMPONENTS
REACTOR COOLING SYSTEMS
REACTORS
SAFETY
WATER COOLED REACTORS
WATER MODERATED REACTORS
220900* - Nuclear Reactor Technology- Reactor Safety
210200 - Power Reactors
Nonbreeding
Light-Water Moderated
Nonboiling Water Cooled