Pipe fracture evaluations for leak-rate detection: Probabilistic models
Creators
- 1. Battelle Memorial Inst., Columbus, OH (United States). Dept. of Engineering Mechanics
Description
This is the second in series of three papers generated from studies on nuclear pipe fracture evaluations for leak-rate detection. This paper focuses on the development of novel probabilistic models for stochastic performance evaluation of degraded nuclear piping systems. It was accomplished here in three distinct stages. First, a statistical analysis was conducted to characterize various input variables for thermo-hydraulic analysis and elastic-plastic fracture mechanics, such as material properties of pipe, crack morphology variables, and location of cracks found in nuclear piping. Second, a new stochastic model was developed to evaluate performance of degraded piping systems. It is based on accurate deterministic models for thermo-hydraulic and fracture mechanics analyses described in the first paper, statistical characterization of various input variables, and state-of-the-art methods of modem structural reliability theory. From this model. the conditional probability of failure as a function of leak-rate detection capability of the piping systems can be predicted. Third, a numerical example was presented to illustrate the proposed model for piping reliability analyses. Results clearly showed that the model provides satisfactory estimates of conditional failure probability with much less computational effort when compared with those obtained from Monte Carlo simulation. The probabilistic model developed in this paper will be applied to various piping in boiling water reactor and pressurized water reactor plants for leak-rate detection applications
Additional details
Publishing Information
- Publisher
- American Society of Mechanical Engineers.
- Imprint Place
- New York, NY (United States)
- ISBN
- 0-7918-0993-5
- Imprint Title
- Creep, fatigue, flaw evaluation, and leak-before-break assessment
- Imprint Pagination
- 301 p.
- Journal Page Range
- p. 255-267.
Conference
- Title
- 1993 pressure vessel and piping conference.
- Dates
- 25-29 Jul 1993.
- Place
- Denver, CO (United States).
INIS
- Country of Publication
- United States
- Country of Input or Organization
- United States
- INIS RN
- 25025918
- Subject category
- S21: SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS; S21: SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS;
- Resource subtype / Literary indicator
- Conference
- Descriptors DEI
- BWR TYPE REACTORS; CRACK PROPAGATION; ELASTICITY; FAILURES; FRACTURE MECHANICS; HEAT TRANSFER; HYDRAULICS; LEAKS; MONTE CARLO METHOD; PIPES; PLASTICITY; PRIMARY COOLANT CIRCUITS; PROBABILITY; PWR TYPE REACTORS; RELIABILITY; STOCHASTIC PROCESSES
- Descriptors DEC
- CALCULATION METHODS; COOLING SYSTEMS; ENERGY TRANSFER; ENRICHED URANIUM REACTORS; MECHANICAL PROPERTIES; MECHANICS; POWER REACTORS; REACTOR COMPONENTS; REACTOR COOLING SYSTEMS; REACTORS; THERMAL REACTORS; WATER COOLED REACTORS; WATER MODERATED REACTORS
Optional Information
- Secondary number(s)
- CONF-930702--Vol.266.