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Unsteady Fluid/Solid Numerical Simulations to Evaluate Thermal Solicitations in PWR Nuclear Plant Mixing Zones

[+] Author Affiliations
Thomas Pasutto, Christophe Péniguel, Marc Sakiz

EDF R&D, Chatou, France

Jean-Michel Stéphan

EDF R&D, Moret-sur-Loing, France

Paper No. PVP2005-71285, pp. 837-844; 8 pages
  • ASME 2005 Pressure Vessels and Piping Conference
  • Volume 4: Fluid Structure Interaction
  • Denver, Colorado, USA, July 17–21, 2005
  • Conference Sponsors: Pressure Vessels and Piping Division
  • ISBN: 0-7918-4189-8 | eISBN: 0-7918-3763-7
  • Copyright © 2005 by ASME


Thermal fatigue of the coolant circuits of PWR plants is a major issue for nuclear safety. The problem is especially accute in mixing zones, like T-junctions, where large differences in water temperature between the two inlets and high levels of turbulence can lead to large temperature fluctuations at the wall. Until recently, studies on the matter had been tackled at EDF using steady methods: the fluid flow was solved with a CFD code using an averaged turbulence model, which led to the knowledge of the mean temperature and temperature variance at each point of the wall. But, being based on averaged quantities, this method could not reproduce the unsteady and 3D effects of the problem, like phase lag in temperature oscillations between two points, which can generate important stresses. Benefiting from advances in computer power and turbulence modelling, a new methodology is now applied, that allows to take these effects into account. The CFD tool Code_Saturne, developped at EDF, is used to solve the fluid flow using an unsteady L.E.S. approach. It is coupled with the thermal code Syrthes, which propagates the temperature fluctuations into the wall thickness. The instantaneous temperature field inside the wall can then be extracted and used for structure mechanics computations (mainly with EDF thermomechanics tool Code_Aster, see joint paper [1]). The purpose of this paper is to present the application of this methodology to the simulation of a straight T-junction mockup, similar to the Residual Heat Remover (RHR) junction found in N4 type PWR nuclear plants, and designed to study thermal striping and cracks propagation. The results are generally in good agreement with the measurements; yet, in certain areas of the flow, progress is still needed in L.E.S. modelling and in the treatment of instantaneous heat transfer at the wall.

Copyright © 2005 by ASME



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