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ABSTRACT
In nuclear safety field, neutronic and thermalhydraulic codes performance is an important issue. New capabilities implementation, as well as models and tools improvements are a significant part of the community effort in looking for better nuclear power plants (NPP) designs. A procedure to analyze the PWR response to local deviations on neutronic or thermalhydraulic parameters is being developed. This procedure includes the simulation of Incore and Excore neutron flux detectors signals. A control rod drop real plant transient is used to validate the used codes and their new capabilities. Cross-section data are obtained by means of the SIMTAB methodology. Detailed thermalhydraulic models were developed: RELAP5 and TRACE models simulate three different azimuthal zones. Besides, TRACE model is performed with a fully three-dimensional core, thus, the cross-flow can be obtained. A Cartesian vessel represents the fuel assemblies and a cylindrical vessel the bypass and downcomer. Simulated detectors signals are obtained and compared with the real data collected during a control rod drop trial at a PWR NPP and also with data obtained with SIMULATE-3K code.
Acknowledgments
The authors would like to acknowledge the economic support provided by Centrales Nucleares Almaraz-Trillo (CNAT) and IBERDROLA Ingeniería y Construcción (Iberinco) for the realization of this work, and express their great appreciation to Arturo López, Juan Antonio Bermejo and Alberto Ortego for their valuable collaboration and their willingness to develop this work. This work has also been supported by the Spanish Ministerio de Economía y Competitividad, through the projects NUC-MULTPHYS (ENE2012-34585) and VALIUN-3D (ENE2011-22823), and the Generalitat Valenciana (GVA), through the project PROMETEO II/2014/008.
Disclosure statement
No potential conflict of interest was reported by the authors.