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Thermal-Hydraulics | |
Nuclear Related Papers |
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THE DETERMINATION OF THE FATIGUE CRACK GROWTH IN REACTOR PRESSURE VESSEL
Darko Korošec, dipl.ing., Slovenian Nuclear Safety Administration Vojkova 59, 1000 Ljubljana, Slovenia Doc. dr. Jelena Vojvodič Gvardjančič, Institute of Metals and Technology, Lepi Pot 11, 1001 Ljubljana, Slovenia Extended Abstract The use of the fracture mechanics in fatigue approach of crack growth in Reactor Pressure Vessel (RPV) is discussed in the present paper. Considering the RPV as one of the most important component in the Pressurized Water Reactor nuclear power plant, the determination of the fatigue crack propagation curve is described. The first research work in this field was already done in Slovenia at the Institute IMK in 1988 and the authors of the present paper intend to continue the work started. The ASME Boiler and Pressure Vessel Code in Section XI (Rules for Inservice Inspection of Nuclear Power Plants Components) requires in certain articles the analytical methods for determination the acceptability of the fatigue crack growth rate in terms of the range of applied stress intensit y factor change ΔKI. This characterisation is governed by the so-called Paris equation. Some different approaches from literature solving this equation are mentioned. This equation involves many different parameters depending on material, operational and other conditions.Service loadings on RPV, their amplitude and frequency for different operational conditions for certain operational period of nuclear power plant are usually known. Using these data the analytical calculation shall be made, taking into account the existing flaws in RPV found during one or more in-service inspections. The component (RPV) containing the flaw(s) is considered to be acceptable for continued service during the next evaluated period only in the case, that the requests from the specific ASME XI articles were satisfied. Analytical evaluation of the flaws and the acceptance criteria are complex, only the approximate solutions are possible. Many of variables and constants needed in the analytical evaluation are not well determined and the correct calculation of the stress and temperature field distribution in the cracked region is not accurate. The evaluation procedures shall be the responsibility of the owner of the nuclear power plant and shall be subject to approval by the regulatory authority (Nuclear Safety Administration). Qualitative crack growth evaluation taking into account the operational history of the RPW is an useful tool for its residual lifetime calculation.
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