Analysis of Pressurized Water Reactor Primary Coolant Leak Events Caused by Thermal Fatigue
We present statistical analyses of pressurized water reactor (PWR) primary coolant leak events caused by thermal fatigue, and discuss their safety significance. Our worldwide data contain 13 leak events (through-wall cracking) in 3509 reactor-years, all in stainless steel piping with diameter less than 25 cm. Several types of data analysis show that the frequency of leak events (events per reactor-year) is increasing with plant age, and the increase is statistically significant. When an exponential trend model is assumed, the leak frequency is estimated to double every 8 years of reactor age, although this result should not be extrapolated to plants much older than 25 years. Difficulties in arresting this increase include lack of quantitative understanding of the phenomena causing thermal fatigue, lack of understanding of crack growth, and difficulty in detecting existing cracks.
- Research Organization:
- Idaho National Lab. (INL), Idaho Falls, ID (United States)
- Sponsoring Organization:
- US Department of Energy (US)
- DOE Contract Number:
- AC07-94ID13223
- OSTI ID:
- 10169
- Report Number(s):
- INEEL/CON-99-00320; TRN: US0103368
- Resource Relation:
- Conference: ESREL '99 - European Safety and Reliability Conference, Munich (DE), 09/13/1999--09/17/1999; Other Information: PBD: 1 Sep 1999
- Country of Publication:
- United States
- Language:
- English
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Related Subjects
22 GENERAL STUDIES OF NUCLEAR REACTORS
36 MATERIALS SCIENCE
FAILURE MODE ANALYSIS
PRIMARY COOLANT CIRCUITS
CRACK PROPAGATION
DATA ANALYSIS
PWR TYPE REACTORS
RELIABILITY
REACTOR SAFETY
STAINLESS STEELS
THERMAL FATIGUE
LEAK EVENTS
PRIMARY COOLANT SYSTEM
THROUGH-WALL CRACKING
PRESSURIZED WATER REACTOR