VISA2. Probability of Reactor Vessel Failure
- Pacific Northwest Lab., Richland, WA (United States)
VISA2 (Vessel Integrity Simulation Analysis) was developed to estimate the failure probability of nuclear reactor pressure vessels under pressurized thermal shock conditions. The deterministic portion of the code performs heat transfer, stress, and fracture mechanics calculations for a vessel subjected to a user-specified temperature and pressure transient. The probabilistic analysis performs a Monte Carlo simulation to estimate the probability of vessel failure. Parameters such as initial crack size and position, copper and nickel content, fluence, and the fracture toughness values for crack initiation and arrest are treated as random variables. Linear elastic fracture mechanics methods are used to model crack initiation and growth. This includes cladding effects in the heat transfer, stress, and fracture mechanics calculations. The simulation procedure treats an entire vessel and recognizes that more than one flaw can exist in a given vessel. The flaw model allows random positioning of the flaw within the vessel wall thickness, and the user can specify either flaw length or length-to-depth aspect ratio for crack initiation and arrest predictions. The flaw size distribution can be adjusted on the basis of different inservice inspection techniques and inspection conditions. The toughness simulation model includes a menu of alternative equations for predicting the shift in the reference temperature of the nil-ductility transition.
- Research Organization:
- Pacific Northwest National Lab. (PNNL), Richland, WA (United States)
- Sponsoring Organization:
- Nuclear Regulatory Commission, Washington, DC (United States). Office of Nuclear Regulatory Research
- OSTI ID:
- 138917
- Report Number(s):
- ESTSC/NRC-000006IPS0200; NESC-1115
- Resource Relation:
- Other Information: PBD: 13 Jan 1992
- Country of Publication:
- United States
- Language:
- English
Similar Records
VISA-II: a computer code for predicting the probability of reactor pressure vessel failure
New techniques for modeling the reliability of reactor pressure vessels