Determination of the critical crack length for steam generator tubing based on fracture-mechanics-based method
Creators
- 1. State Key Laboratory of Multiphase Flow in Power Engineering, Xi'an Jiaotong University, Xi'an 710049 (China)
- 2. School of Energy and Power Engineering, Xi'an Jiaotong University, Xi'an 710049 (China)
Description
Highlights: → The critical crack lengths at different stages are determined in the form of graphs based on fracture-mechanics approach. → The crack propagation sequence and mechanism were determined. → The conditions for reasonable extrapolations and unreasonable extrapolations during the accelerated life test were identified. → These messages are vital to predict the life of SG tubing correctly. - Abstract: The steam generator (SG) tubing, as a key ingredient in the primary coolant circuit, is the weakest link that affects the availability and safety of a nuclear power plant. For safety reasons, it is very important to predict the life span of SG tubing. The critical crack lengths at different stages should be determined when the life span can be predicted. In this article, the critical crack lengths at different stages are determined in the form of graphs based on fracture-mechanics approach. From these graphs, it is concluded that the critical crack lengths for fatigue, stress corrosion cracking (SCC), and rupture are 0.11 mm, 0.36 mm and 7.20 mm respectively under accident conditions, and 1 mm, 2 mm, 43 mm respectively under normal conditions. It can also be concluded that the crack propagation mechanism can be divided successively into three or four stages, namely the corrosion stage, the fatigue stage (if the load is turbulent), the SCC stage and the rupture stage. Finally, some advices for the accelerated life test are given.
Availability note (English)
Available from http://dx.doi.org/10.1016/j.anucene.2011.05.009Additional details
Identifiers
- DOI
- 10.1016/j.anucene.2011.05.009;
- PII
- S0306-4549(11)00198-8;
Publishing Information
- Journal Title
- Annals of Nuclear Energy (Oxford)
- Journal Volume
- 38
- Journal Issue
- 9
- Journal Page Range
- p. 1900-1905
- ISSN
- 0306-4549
- CODEN
- ANENDJ
INIS
- Country of Publication
- United Kingdom
- Country of Input or Organization
- International Atomic Energy Agency (IAEA)
- INIS RN
- 43120407
- Subject category
- S42: ENGINEERING; S22: GENERAL STUDIES OF NUCLEAR REACTORS;
- Descriptors DEI
- ACCIDENTS; AVAILABILITY; CRACK PROPAGATION; CRACKS; EXTRAPOLATION; FATIGUE; FRACTURE MECHANICS; NUCLEAR POWER PLANTS; PRIMARY COOLANT CIRCUITS; REACTOR SAFETY; RUPTURES; STEAM GENERATORS; STRESS CORROSION
- Descriptors DEC
- BOILERS; CHEMICAL REACTIONS; COOLING SYSTEMS; CORROSION; ENERGY SYSTEMS; FAILURES; MATHEMATICAL SOLUTIONS; MECHANICAL PROPERTIES; MECHANICS; NUCLEAR FACILITIES; NUMERICAL SOLUTION; POWER PLANTS; REACTOR COMPONENTS; REACTOR COOLING SYSTEMS; SAFETY; THERMAL POWER PLANTS; VAPOR GENERATORS
Optional Information
- Copyright
- Copyright (c) 2011 Elsevier Science B.V., Amsterdam, The Netherlands, All rights reserved.