During the last several years a considerable effort was devoted and progress has been made in various countries and organizations in incorporating full three-dimensional (3D) reactor core models into system transient codes, it allows performing of a “best-estimate” calculation of interactions between the core behavior and plant dynamics. Several benchmarks have been developed to verify and validate the capability of the coupled codes in order to analyze complex transients with coupled core-plant interactions for different types of reactors. In December 2008 the NEA/OECD Nuclear Science Committee (NSC) Bureau has expressed support for the coupled Kalinin-3 benchmark problem in general to become an international standard problem for validation of the best-estimate safety codes. This benchmark defines a coupled code problem for further validation of thermal-hydraulics system codes for application to Russian-designed VVER-1000 reactors based on actual plant data from the Russian NPP Kalinin Unit #3 (Kalinin-3). The selected transient “Switching-off of one Main Circulation Pump (MCP)” is performed at a nominal power and leads to asymmetric core conditions with broad ranges of the parameter changes. The available real plant experimental data is very well documented and made these benchmark problems very valuable. Measurements were carried out with a quite high frequency and their uncertainties are known for almost all measured parameters. This fact allows applying the studied transient not only for validation purposes but also for uncertainty analysis as a part of the NEA/OECD LWR Uncertainty Analysis in Modeling (UAM) Benchmark. The purpose of this benchmark is four-fold: to verify the capability of system codes to analyze complex transients with coupled core-plant interactions and complicated fluid mixing phenomena, to fully test the 3D neutronics/thermal-hydraulic coupling, to evaluate discrepancies between predictions of the coupled codes in best-estimate transient simulations with measured data, to perform uncertainty analysis having at disposal not only the measured values but also their accuracy. This report provides the calculations for first three points.
GRNSPG/UNIPI Activities on KALININ-3 Coolant Transient Benchmark: Best Estimate Coupled Plant Transient Modeling
D'AURIA, FRANCESCO SAVERIO
2013-01-01
Abstract
During the last several years a considerable effort was devoted and progress has been made in various countries and organizations in incorporating full three-dimensional (3D) reactor core models into system transient codes, it allows performing of a “best-estimate” calculation of interactions between the core behavior and plant dynamics. Several benchmarks have been developed to verify and validate the capability of the coupled codes in order to analyze complex transients with coupled core-plant interactions for different types of reactors. In December 2008 the NEA/OECD Nuclear Science Committee (NSC) Bureau has expressed support for the coupled Kalinin-3 benchmark problem in general to become an international standard problem for validation of the best-estimate safety codes. This benchmark defines a coupled code problem for further validation of thermal-hydraulics system codes for application to Russian-designed VVER-1000 reactors based on actual plant data from the Russian NPP Kalinin Unit #3 (Kalinin-3). The selected transient “Switching-off of one Main Circulation Pump (MCP)” is performed at a nominal power and leads to asymmetric core conditions with broad ranges of the parameter changes. The available real plant experimental data is very well documented and made these benchmark problems very valuable. Measurements were carried out with a quite high frequency and their uncertainties are known for almost all measured parameters. This fact allows applying the studied transient not only for validation purposes but also for uncertainty analysis as a part of the NEA/OECD LWR Uncertainty Analysis in Modeling (UAM) Benchmark. The purpose of this benchmark is four-fold: to verify the capability of system codes to analyze complex transients with coupled core-plant interactions and complicated fluid mixing phenomena, to fully test the 3D neutronics/thermal-hydraulic coupling, to evaluate discrepancies between predictions of the coupled codes in best-estimate transient simulations with measured data, to perform uncertainty analysis having at disposal not only the measured values but also their accuracy. This report provides the calculations for first three points.I documenti in IRIS sono protetti da copyright e tutti i diritti sono riservati, salvo diversa indicazione.
