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 behaviour of cracks under PTS loading conditions due to the emergency cooling during PTS transient like SBLOCA. This paper explains the Research and Development program started at Electricite´ De France about the cooling phenomena of a PWR vessel after a Pressurised Thermal Shock. The numerical results are obtained with the E.D.F ThermalHydraulic code (Code_Saturne) coupled with the thermal-solid code SYRTHES to take into account the conjugate heat transfer on the cooling of the vessel. We first explain the global methodology with a progress report on the state of the art of the tools available to simulate the different scenari displayed within the frame of the plant life project in order to reassess the integrity of the RPV, taking into account the evolution of some input data, such as the new value of end of life (EOL) fluence, the feedback results of surveillance program and the evolution of the functional requirements. The main results are presented and are related to the evaluation of the RPV integrity during a Small Break Loss Of Coolant Accident transient for 900 and 1300 MWe nuclear plant. On the whole, the main purpose of the numerical CFD 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 margins. In a second time, a new analysis is performed to assess an accurate temperature distribution in the RPV. Indeed, from a physical phenomena point of view, the EDF thermalhydraulic tool Code_Saturne is now qualified in order to assess single phase transient but in the case where the cold legs are partially filled with steam, it becomes a two-phase problem and new important effects occur, such as condensation due to the emergency core cooling injections of sub-cooled water. Thus, an advanced prediction of RPV thermal loading during these transients requires sophisticated two-phase, local scale, 3D codes. In that purpose, a program has been set up to extend the capabilities of the Neptune_CFD two-phase solver which is the tool able to solve two phase flow configuration. In a same time, A simplified approach has showed that for a type of transient weakly uncovered, a free surface calculation was sufficient to respect the necessary criteria of safety. A Qualification study was carried out on the Hybiscus experimental E.D.F facility, representing a cold leg with ECC injection and a third down comer. Temperature profiles have been compared and are presented and analysed here, showing encouraging results.
Use of a CFD-Tool for Assessment of the Reactor Pressure Vessel Integrity in Pressurised Thermal Shock Conditions: Thermal-Hydraulic Studies of a Safety Injection in a PWR Plant — Methodology and Present Limitation
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Martin, A, Bosse, S, & Lestang, F. "Use of a CFD-Tool for Assessment of the Reactor Pressure Vessel Integrity in Pressurised Thermal Shock Conditions: Thermal-Hydraulic Studies of a Safety Injection in a PWR Plant — Methodology and Present Limitation." Proceedings of the ASME 2007 Pressure Vessels and Piping Conference. Volume 7: Operations, Applications and Components. San Antonio, Texas, USA. July 22–26, 2007. pp. 45-56. ASME. https://doi.org/10.1115/PVP2007-26316
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