Integrity evaluation methods for nuclear Reactor Pressure Vessels (RPVs) under Pressurised Thermal Shock (PTS) loading are applied by French Utility. They are based on the analysis of the behavior of relatively shallow cracks under loading PTS conditions due to the emergency cooling during SBLOCA (Small Break Loss of Coolant Accident) transients. This paper presents the Research and Development program started at E.D.F on the Computational Fluid Dynamic (CFD) determination of the cooling phenomena of a PWR vessel during a Pressurised Thermal Shock. The numerical results are obtained with the thermal-hydraulic tool Code_Saturne, in combination with the thermal-solid code SYRTHES to take into account the coupled effect of heat transfer between the fluid flow and the vessel. Based on the global and local Thermal-hydraulic analysis of a Small Break Loss of Coolant Accident transient, the paper presents mainly a parametric study whose purpose is to understand the main phenomena which can lead to better estimating the margin factors. The geometry studied represents a third of a PWR pressure vessel and the configuration investigated is related to the injection of cold water in the vessel during a SBLOCA transient. Conservative initial and boundary conditions for the CFD calculation are derived from the global Thermal-hydraulic analysis. Both the fluid behavior and its impact on the solid part formed by cladding and base metal are considered. On the whole, the main purpose of the numerical thermal-hydraulic studies is to accurately estimate the distribution of fluid temperature in the down comer and the heat transfer coefficients on the inner RPV surface for a fracture mechanics computation which will subsequently assess the associated RPV safety margin factors.
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ASME 2005 Pressure Vessels and Piping Conference
July 17–21, 2005
Denver, Colorado, USA
Conference Sponsors:
- Pressure Vessels and Piping Division
ISBN:
0-7918-4192-8
PROCEEDINGS PAPER
CFD-Tool for Assessment of the Reactor Pressure Vessel Integrity in Pressure Thermal Shock Conditions: Influence of Temperature Safety Injection and Fluid-Structure Thermal Coupling Available to Purchase
A. Martin,
A. Martin
Electricite´ de France, Chatou, France
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D. Monfort,
D. Monfort
Electricite´ de France, Chatou, France
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G. Bezdikian,
G. Bezdikian
Electricite´ de France, Saint-Denis Cedex, France
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F. Beaud,
F. Beaud
Electricite´ de France, Villeurbanne Cedex, France
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F. Lestang,
F. Lestang
Electricite´ de France, Villeurbanne Cedex, France
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C. Vit
C. Vit
Incka SA, Boulogne-Billancourt
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A. Martin
Electricite´ de France, Chatou, France
D. Monfort
Electricite´ de France, Chatou, France
G. Bezdikian
Electricite´ de France, Saint-Denis Cedex, France
F. Beaud
Electricite´ de France, Villeurbanne Cedex, France
F. Lestang
Electricite´ de France, Villeurbanne Cedex, France
C. Vit
Incka SA, Boulogne-Billancourt
Paper No:
PVP2005-71585, pp. 33-40; 8 pages
Published Online:
July 29, 2008
Citation
Martin, A, Monfort, D, Bezdikian, G, Beaud, F, Lestang, F, & Vit, C. "CFD-Tool for Assessment of the Reactor Pressure Vessel Integrity in Pressure Thermal Shock Conditions: Influence of Temperature Safety Injection and Fluid-Structure Thermal Coupling." Proceedings of the ASME 2005 Pressure Vessels and Piping Conference. Volume 7: Operations, Applications, and Components. Denver, Colorado, USA. July 17–21, 2005. pp. 33-40. ASME. https://doi.org/10.1115/PVP2005-71585
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