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Science and Technology of Nuclear Installations
Volume 2009 (2009), Article ID 835162, 7 pages
http://dx.doi.org/10.1155/2009/835162
Research Article

CFD Simulation of Thermal-Hydraulic Benchmark V1000CT-2 Using ANSYS CFX

Forschungszentrum Dresden-Rossendorf (FZD), Institute of Safety Research, P.O. Box 510119, 01314 Dresden, Germany

Received 16 July 2008; Revised 20 October 2008; Accepted 11 January 2009

Academic Editor: Bousbia Salah Anis

Copyright © 2009 Thomas Höhne. This is an open access article distributed under the Creative Commons Attribution License, which permits unrestricted use, distribution, and reproduction in any medium, provided the original work is properly cited.

Abstract

Plant measured data from VVER-1000 coolant mixing experiments were used within the OECD/NEA and AER coupled code benchmarks for light water reactors to test and validate computational fluid dynamic (CFD) codes. The task is to compare the various calculations with measured data, using specified boundary conditions and core power distributions. The experiments, which are provided for CFD validation, include single loop cooling down or heating-up by disturbing the heat transfer in the steam generator through the steam valves at low reactor power and with all main coolant pumps in operation. CFD calculations have been performed using a numerical grid model of 4.7 million tetrahedral elements. The Best Practice Guidelines in using CFD in nuclear reactor safety applications has been used. Different advanced turbulence models were utilized in the numerical simulation. The results show a clear sector formation of the affected loop at the downcomer, lower plenum and core inlet, which corresponds to the measured values. The maximum local values of the relative temperature rise in the calculation are in the same range of the experiment. Due to this result, it is now possible to improve the mixing models which are usually used in system codes.