CFD Analysis of Coolant Mixing in VVER-1000 Pressure Vessel
16th Symposium of AER on VVER Reactor Physics and Reactor Safety (2006, Bratislava, Slovakia)
Reactor Dynamics, Thermal Hydraulics and Safety Analysis
In the framework of the VVER-1000 Coolant Transient (V1000CT) Benchmarks coolant mixing and main steam line break (MSLB) benchmarks for VVER-1000 reactors were defined. Reference plant is Kozloduy-6, measured data from NPP mixing experiments are available to validate the calculation results, and – within the benchmark program – code-to-code comparison is to be used to evaluate the reactor vessel thermal hydraulic models. As a part of this program an exercise is specified to test the capability of vessel thermal-hydraulic models to represent the vessel mixing. Plant measured data from VVER-1000/V320 coolant mixing experiments are available to test and validate vessel mixing models (CFD, coarse-mesh and mixing matrix). The task of the benchmarks is to compare the calculations with measured data, using specified vessel boundary conditions and core power distribution. The measurements were conducted during the plant commissioning phase at Kozloduy-6 and Kalinin-1, 2. CAD geometry model of the reactor pressure vessel is provided as a part of the source information for the plant data.
In the present paper mixing calculations based on the V1000CT benchmark definition will be presented. For the calculations the ANSYS CFX-10 three-dimensional Computational Fluid Dynamics code was used. The calculations cover steady state analyses of the final state of the Kozloduy-6 mixing test. In this state on cold leg No. 1. the coolant temperature is higher (282.2 °C) than the temperature of other cold legs (cold leg No. 2.: 269.9 °C, No. 3.: 269 °C and No. 4.: 269.2 °C). Measured fuel assembly outlet temperatures, fuel assembly mixing coefficients and estimated fuel assembly inlet temperatures are provided. For the modeling of the VVER-1000 reactor pressure vessel non-structured tetrahedral mesh was used. The mesh was generated by the code ICEM using the provided CAD geometry of the reactor pressure vessel. The results of the CFX calculations will be compared with the available measured and estimated plant data.