%0 Journal Article
%T Validation and Application of the Thermal Hydraulic System Code TRACE for Analysis of BWR Transients
%A V. H. S¨˘nchez
%A M. Thieme
%A W. Tietsch
%J Science and Technology of Nuclear Installations
%D 2012
%I Hindawi Publishing Corporation
%R 10.1155/2012/247482
%X 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. 1. Introduction The use of validated numerical simulation tools for the analysis of the plant response under off normal conditions is mandatory. In the framework of the Code Applications and Maintenance Program (CAMP) of the US NRC, the TRACE code is being validated for LWR safety investigations [1]. An extensive validation of coupled neutron kinetics/thermal hydraulic codes is taking place worldwide in the frame of national and international benchmarks related to both pressurized and boiling water reactors (PWR and BWRs) using plant data such as the Peach Bottom turbine trip test [2, 3], the Oskarshamn-2 Instability event [4] and the Ringhals Loss of Feedwater Case [5]. The Karlsruhe Institute of Technology (KIT) is participating on CAMP program and performing validation work for light water reactors (LWRs) [6] and Generation 4 reactors [7]. For the validation of safety-relevant TRACE heat transfer models, data from different bundle tests such as the NUPEC BFBT and PSBT (PWR Subchannel and Bundle Test) tests are available from the international OECD/NEA Benchmarks [8, 9]. Important data for the validation of models related to single and two-phase flow pressure drop, void fraction, burnout, and DNB are
%U http://www.hindawi.com/journals/stni/2012/247482/