Probabilistic assessment of decoupling loss-of-coolant accident and earthquake in nuclear power plant design
This paper describes a research project conducted at Lawrence Livermore National Laboratory to establish a technical basis for reassessing the requirement of combining large loss-of-coolant-accident (LOCA) and earthquake loads in nuclear power plant design. A large LOCA is defined herein as a double-ended guillotine break of the primary reactor coolant loop piping (the hot leg, cold leg, and crossover) of a pressureized water reactor (PWR). A systematic probability approach has been employed to estimate the probability of a large LOCA directly and indirectly induced by earthquakes. The probability of a LOCA directly induced by earthquakes was assessed by a numerical simulation of pipe rupture of a reactor coolant system. The simulation employed a deterministic fracture mechanics model which dictates the fatigue growth of pre-existing cracks in the pipe. The simulation accounts for the stochastic nature of input elements such as the initial crack size distribution, the crack occurrence rate, crack and leak detection probabilities as functions of crack size, plant transient occurrence rates, the seismic hazard, stress histories, and crack growth model parameters. Effects on final results due to variation an uncertainty of input elements were assessed by a limited sensitivity study. Results of the simulation indicate that the probability of a double-ended guillotine break, either with or without an earthquake, is very small (on the orer of 10/sup -12/). The probability of a leak was found to be several orders of magnitudes greater than that of a complete break.
- Research Organization:
- Lawrence Livermore National Lab., CA (USA); Science Applications, Inc., Palo Alto, CA (USA)
- DOE Contract Number:
- W-7405-ENG-48
- OSTI ID:
- 6470594
- Report Number(s):
- UCRL-84186; CONF-810801-25; ON: TI85016681
- Country of Publication:
- United States
- Language:
- English
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210200 -- Power Reactors
Nonbreeding
Light-Water Moderated
Nonboiling Water Cooled
22 GENERAL STUDIES OF NUCLEAR REACTORS
220900* -- Nuclear Reactor Technology-- Reactor Safety
ACCIDENTS
COOLING SYSTEMS
ENERGY SYSTEMS
ENERGY TRANSFER
FAILURES
FLUID MECHANICS
HEAT TRANSFER
HYDRAULICS
LOSS OF COOLANT
MATERIALS
MATHEMATICAL MODELS
MECHANICAL PROPERTIES
MECHANICS
PIPES
PRIMARY COOLANT CIRCUITS
PROBABILITY
PWR TYPE REACTORS
REACTOR ACCIDENTS
REACTOR COMPONENTS
REACTOR COOLING SYSTEMS
REACTOR MATERIALS
REACTOR SAFETY
REACTORS
RISK ASSESSMENT
SAFETY
SEISMIC EFFECTS
SIMULATION
STRESSES
ULTIMATE STRENGTH
WATER COOLED REACTORS
WATER MODERATED REACTORS