Interim Reliability Evaluation Program: analysis of the Browns Ferry, Unit 1, nuclear plant. Main report
A probabilistic risk assessment (PRA) was made of the Browns Ferry, Unit 1, nuclear plant as part of the Nuclear Regulatory Commission's Interim Reliability Evaluation Program (IREP). Specific goals of the study were to identify the dominant contributors to core melt, develop a foundation for more extensive use of PRA methods, expand the cadre of experienced PRA practitioners, and apply procedures for extension of IREP analyses to other domestic light water reactors. Event tree and fault tree analyses were used to estimate the frequency of accident sequences initiated by transients and loss of coolant accidents. External events such as floods, fires, earthquakes, and sabotage were beyond the scope of this study and were, therefore, excluded. From these sequences, the dominant contributors to probable core melt frequency were chosen. Uncertainty and sensitivity analyses were performed on these sequences to better understand the limitations associated with the estimated sequence frequencies. Dominant sequences were grouped according to common containment failure modes and corresponding release categories on the basis of comparison with analyses of similar designs rather than on the basis of detailed plant-specific calculations.
- Research Organization:
- EG and G Idaho, Inc., Idaho Falls (USA); Energy, Inc., Seattle, WA (USA)
- DOE Contract Number:
- AC07-76ID01570
- OSTI ID:
- 5047497
- Report Number(s):
- NUREG/CR-2802; EGG-2199; ON: DE82020539
- Country of Publication:
- United States
- Language:
- English
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21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
BROWNS FERRY-1 REACTOR
MELTDOWN
RISK ASSESSMENT
FAILURE MODE ANALYSIS
FAULT TREE ANALYSIS
REACTOR SAFETY
RELIABILITY
ACCIDENTS
BWR TYPE REACTORS
ENRICHED URANIUM REACTORS
MIXED SPECTRUM REACTORS
POWER REACTORS
REACTOR ACCIDENTS
REACTORS
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SYSTEM FAILURE ANALYSIS
SYSTEMS ANALYSIS
WATER COOLED REACTORS
WATER MODERATED REACTORS
220900* - Nuclear Reactor Technology- Reactor Safety
210100 - Power Reactors
Nonbreeding
Light-Water Moderated
Boiling Water Cooled