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Science and Technology of Nuclear Installations
Volume 2012 (2012), Article ID 247482, 9 pages
Research Article

Validation and Application of the Thermal Hydraulic System Code TRACE for Analysis of BWR Transients

1Institute for Neutron Physics and Reactor Technology (INR), Karlsruhe Institute of Technology (KIT), 76344 Eggenstein-Leopoldshafen, Germany
2TÜV SÜD Industrie Service GmbH, Westendstraße 199, 80686 Munich, Germany
3Westinghouse Electric Germany GmbH, 68140 Mannheim, Germany

Received 13 April 2012; Accepted 7 August 2012

Academic Editor: Boštjan Končar

Copyright © 2012 V. H. Sánchez et al. 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.


The Karlsruhe Institute of Technology (KIT) is participating on (Code Applications and Maintenance Program) CAMP of the US Nuclear Regulatory Commission (NRC) to validate TRACE code for LWR transient analysis. The application of TRACE for the safety assessment of BWR requires a throughout verification and validation using experimental data from separate effect and integral tests but also using plant data. The validation process is normally focused on safety-relevant phenomena for example, pressure drop, void fraction, heat transfer, and critical power models. The purpose of this paper is to validate selected BWR-relevant TRACE-models using both data of bundle tests such as the (Boiling Water Reactor Full-Size Fine-Mesh Bundle Test) BFBT and plant data recorded during a turbine trip event (TUSA) occurred in a Type-72 German BWR plant. For the validation, TRACE models of the BFBT bundle and of the BWR plant were developed. The performed investigations have shown that the TRACE code is appropriate to describe main BWR-safety-relevant phenomena (pressure drop, void fraction, and critical power) with acceptable accuracy. The comparison of the predicted global BWR plant parameters for the TUSA event with the measured plant data indicates that the code predictions are following the main trends of the measured parameters such as dome pressure and reactor power.