Probability of pipe fracture in the primary coolant loop of a PWR plant. Volume 1. Summary, Load Combination Program. Project I final report

Technical Report · Mon Jun 01 04:00:00 EDT 1981

DOI:https://doi.org/10.2172/6469612· OSTI ID:6469612

Lu, S.; Streit, R. D.; Chou, C. K.

This report summarizes work performed to establish a technical basis for the NRC to use in reassessing its requirement that earthquake and large loss-of-coolant accident (LOCA) loads be combined in the design of nuclear power plants. A systematic probabilistic approach is used to treat the random nature of earthquake and transient loading and to estimate the probability of large LOCAs that are directly and indirectly induced by earthquakes. A large LOCA is defined in this report as a double-ended guillotine break of the primary reactor coolant loop piping (the hot leg, cold leg, and crossover) of a pressurized water reactor (PWR). Unit 1 of the Zion Nuclear Power Plant, a four-loop PWR, is the demonstration plant used in this study. To estimate the probability of a large LOCA directly induced by earthquakes, only fatigue crack growth resulting from the combined effects of thermal, pressure, seismic, and other cyclic loads is considered. Fatigue crack growth is simulated by a deterministic fracture mechanics model with stochastic inputs of initial crack size distribution, material properties, stress histories, and leak detection probability. Results of the simulation indicate that the probability of a double-ended guillotine break, either with or without earthquake, is very small (on the order of 10/sup -12/). The probability of a leak was found to be several orders of magnitude greater than that of a large LOCA, complete pipe rupture. A limited investigation involving engineering judgment of a double-ended guillotine break indirectly induced by an earthquake is also reported.

• Research Organization: Lawrence Livermore National Lab., CA (USA)
• DOE Contract Number: W-7405-ENG-48
• OSTI ID: 6469612
• Report Number(s): NUREG/CR-2189-Vol.1; UCID-18967-Vol.1; ON: TI84017628
• Country of Publication: United States
• Language: English

Related Subjects

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
210200 -- Power Reactors
Nonbreeding
Light-Water Moderated
Nonboiling Water Cooled
22 GENERAL STUDIES OF NUCLEAR REACTORS
220900* -- Nuclear Reactor Technology-- Reactor Safety
ACCIDENTS
CRACK PROPAGATION
EARTHQUAKES
ENRICHED URANIUM REACTORS
FAILURES
FATIGUE
FRACTURES
LEAKS
LOSS OF COOLANT
MECHANICAL PROPERTIES
PIPES
POWER REACTORS
PWR TYPE REACTORS
REACTOR ACCIDENTS
REACTORS
SEISMIC EFFECTS
SEISMIC EVENTS
THERMAL REACTORS
WATER COOLED REACTORS
WATER MODERATED REACTORS
ZION-1 REACTOR