In the case of a severe accident scenario of a pressurized water reactor which includes cracking of the vessel bottom head, it is crucial to predict the leak rate and hence the crack size for the ex-vessel accident management. We present an experimental framework to analyze the crack propagation under such severe conditions for different 16MND5 French nuclear steel grades. An original experimental setup has been designed in order to perform bi-axial tests (tensile load independent of internal pressure) at high temperatures (1180K – 1280K) on tubular test specimens. The temperature loading and the mechanical loading can be set to reproduce the stress distribution of the hemispherical vessel bottom head submitted to an internal pressure. Moreover, the test was designed to be easily transposable to the real structure in terms of crack propagation and depressurization thanks to an energy based scaling methodology. We observed the crack initiation and propagation with two high speed digital cameras. Force, internal pressure, displacement and temperature fields were also measured and synchronized with the optical measurements. The different creep stages are observed and characterized. The crack propagation and opening history have been measured. During crack initiation and propagation stages, the depressurization can be correlated with the crack geometry. Finally, the setup has been designed in order to validate future numerical analysis.
- Pressure Vessels and Piping
Unstable Crack Propagation Under Severe Accident Scenario Conditions in a Pressurized Water Reactor
Tardif, N, Coret, M, & Combesure, A. "Unstable Crack Propagation Under Severe Accident Scenario Conditions in a Pressurized Water Reactor." Proceedings of the ASME 2009 Pressure Vessels and Piping Conference. Volume 5: High Pressure Technology; Nondestructive Evaluation Division; Student Paper Competition. Prague, Czech Republic. July 26–30, 2009. pp. 343-350. ASME. https://doi.org/10.1115/PVP2009-77258
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