Nuclear power plants piping systems necessarily include connections of branches conveying fluids at different temperatures. Thermal-hydraulic fluctuations arising from the turbulent mixing of the flows possibly affect the inner wall of the pipes and lead to fatigue damage. This paper aims at presenting an analytical model for the mixing zone thermal-mechanical problem and consecutive crack initiation and propagation analyses.
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 relatively shallow cracks under loading PTS conditions due to the emergency cooling during SBLOCA transients. This paper presents the Research and Development program started at E.D.F on the CFD determination of the cooling phenomena of a PWR vessel during a Pressurised Thermal Shock. The numerical results are obtained with the thermalhydraulic tools N3S and 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. We first explain the recent improvement of the thermalhydraulic analysis with the global definition of the SBLOCA transient and the local analysis in the downcomer. Then, the qualification task of the EDF numerical tools is described. In order to reach this purpose, we have investigated several configurations related to an injection of cold water and focused our analysis particularly in the cold leg but also in a the downcomer. Two experiment test cases have been studied. A comparison between experiment and numerical results in terms of temperature field is presented. On the whole, the main purpose of the numerical thermalhydraulic studies is to accurately estimate the distribution of fluid temperature in the downcomer and the heat transfer coefficients on the inner RPV surface for a fracture mechanics computation which will subsequently assess the associated RPV safety margins.
Within the framework of the nuclear power plant lifetime issue, the assessment of the French 900 MWe (3-loops) series Reactor Pressure Vessel (RPV) integrity was performed. A simplified analysis has shown that one of the most severe loading condition is given by the small break loss of coolant accidents (SBLOCA) due to the pressurized injection of cold water (9°C) into the cold leg and down comer of the RPV. Two main physical phenomena, considered important for the RPV cooling transient, were identified during numerical application obtained with EDF CFD tools. These phenomena are the fluid flow separation and the plume oscillations in the down comer. In order to consolidate these numerical results with the EDF home code, called Code_Saturne, an experimental study has been carried out with the new EDF R&D facility. This transparent experimental model is based on the representation at 1/2 scale of a cold leg and a third of down comer including a thermal shield. The experiments were realized by injecting of salt water flow (density effects) in the cold leg according to a similitude study based on Froude number conservation between experiments and reactor scenarios. Firstly, this paper presents the main qualitative experimental results, based essentially on visualizations of different injections of dyed salt water in the cold leg and in the down comer. The physical phenomena observed showed a qualitative good agreement between visualizations and numerical results. Secondly, this paper presents the first experimental results of the assessment of the fluid flow separation in the experimental model obtained with temperature probes inserted in the down comer. We showed, in the experiments analysis, the fluid flow separation and the jet oscillations were detected. The next step will consist to compare these quantitative experiments with numerical study which will be carry out with Code_Saturne.
This paper explains the Research and Development program started at E.D.F about the cooling phenomena of a PWR vessel after a Pressurized Thermal Shock. The numerical results are obtained with the 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. The geometry used represents a four loop PWR plant. In this calculation, the simulated geometry takes into account as much as possible the exact geometry of the lower plenum such as its columns and plates instrumentation. The configuration investigated is related to the injection of cold water in the vessel during a penalizing operating transient and its impact on the solid part formed by cladding and base metal. Numerical results are given in terms of temperature field in the cold legs and in the down comer. The obtained numerical description of the transient (internal pressure and temperature field within the vessel) is used as boundary conditions for a full mechanical computation of the stresses. This thermal–mechanical transient is obtained by F.E. simulation using the F.E. code Code_Aster on a 3D mesh of the vessel, covering the two core–shells and their circumferential welds, as well as the internal cladding. Based on an analytical method specially established for underclad flaws, the corrected stress intensity factor Kβ during the transient is evaluated for an hypothetical flaw, by extracting the stresses along a radial segment. The severity of the flaw with respect to the transient is quantified by the minimum of the ratio KIc/Kβ, where KIc refers to the base metal fracture toughness for brittle initiation. The evolution of the severity with the position of the hypothetical flaw is studied and compared with the results given by a classical uni–dimensional method. The results show that such a complete thermal–hydraulic and mechanical 3–dimensional analysis allows to reduce considerably the severity of the flaws, thus improving the integrity of the RPV.
Integrity evaluation methods for nuclear reactor pressure vessels (RPVs) under pressurized 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 small break loss of coolant accident (SBLOCA) transients. This paper presents the research and development program started at EDF on the computational fluid dynamics (CFD) determination of the cooling phenomena of a PWR vessel during a pressurized thermal shock. The numerical results are obtained with the thermal-hydraulic tool Code̱Saturne, 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, this paper presents mainly a parametric study that helps to understand the main phenomena that 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. The main purpose of the numerical thermal-hydraulic studies is to accurately estimate the distribution of fluid temperature in the downcomer 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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