Reactor pressure vessel failure probability following through-wall cracks due to pressurized thermal shock events
A fracture mechanics model was developed at the Pacific Northwest Laboratory (PNL) to predict the behavior of a reactor pressure vessel following a through-wall crack that occurs during a pressurized thermal shock (PTS) event. This study, which contributed to a US Nuclear Regulatory Commission (NRC) program to study PTS risk, was coordinated with the Integrated Pressurized Thermal Shock (IPTS) Program at Oak Ridge National Laboratory (ORNL). The PNL fracture mechanics model uses the critical transients and probabilities of through-wall cracks from the IPTS Program. The PNL model predicts the arrest, reinitiation, and direction of crack growth for a postulated through-wall crack and thereby predicts the mode of vessel failure. A Monte-Carlo type of computer code was written to predict the probabilities of the alternative failure modes. This code treats the fracture mechanics properties of the various welds and plates of a vessel as random variables. Plant-specific calculations were performed for the Oconee-1, Calvert Cliffs-1, and H.B. Robinson-2 reactor pressure vessels for the conditions of postulated transients. The model predicted that 50% or more of the through-wall axial cracks will turn to follow a circumferential weld. The predicted failure mode is a complete circumferential fracture of the vessel, which results in a potential vertically directed missile consisting of the upper head assembly. Missile arrest calculations for the three nuclear plants predict that such vertical missiles, as well as all potential horizontally directed fragmentation type missiles, will be confined to the vessel enclosre cavity. The PNL failure mode model is recommended for use in future evaluations of other plants, to determine the failure modes that are most probable for postulated PTS events.
- Research Organization:
- Pacific Northwest Lab., Richland, WA (USA)
- DOE Contract Number:
- AC06-76RL01830
- OSTI ID:
- 5777623
- Report Number(s):
- NUREG/CR-4483; PNL-5727; ON: TI86010328
- Country of Publication:
- United States
- Language:
- English
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CALVERT CLIFFS-1 REACTOR
COMPUTERIZED SIMULATION
CONTAINERS
CRACK PROPAGATION
CRACKS
ENRICHED URANIUM REACTORS
EVALUATION
FABRICATION
FAILURE MODE ANALYSIS
FAILURES
FRACTURE MECHANICS
JOINING
MATHEMATICAL MODELS
MECHANICS
MONTE CARLO METHOD
OCONEE-1 REACTOR
OPERATION
PHYSICAL RADIATION EFFECTS
POWER REACTORS
PRESSURE VESSELS
PROBABILITY
PWR TYPE REACTORS
RADIATION EFFECTS
REACTOR OPERATION
REACTORS
ROBINSON-2 REACTOR
SIMULATION
SYSTEM FAILURE ANALYSIS
SYSTEMS ANALYSIS
THERMAL SHOCK
TRANSIENTS
WATER COOLED REACTORS
WATER MODERATED REACTORS
WELDING