CFD Simulation of Thermal‐Hydraulic Benchmark V1000CT‐2 Using ANSYS CFX

Thomas Höhne

doi:10.1155/2009/835162

Abstract

Plant measured data from VVER‐1000 coolant mixing experiments were used within the OECD/NEA and AER coupled code benchmarks for light water reactors to test and validate computational fluid dynamic (CFD) codes. The task is to compare the various calculations with measured data, using specified boundary conditions and core power distributions. The experiments, which are provided for CFD validation, include single loop cooling down or heating‐up by disturbing the heat transfer in the steam generator through the steam valves at low reactor power and with all main coolant pumps in operation. CFD calculations have been performed using a numerical grid model of 4.7 million tetrahedral elements. The Best Practice Guidelines in using CFD in nuclear reactor safety applications has been used. Different advanced turbulence models were utilized in the numerical simulation. The results show a clear sector formation of the affected loop at the downcomer, lower plenum and core inlet, which corresponds to the measured values. The maximum local values of the relative temperature rise in the calculation are in the same range of the experiment. Due to this result, it is now possible to improve the mixing models which are usually used in system codes.

Highlights

Several mixing phenomena characterize the various operating conditions of pressurized water reactors (PWRs) and influence the safety analyses of the plant operating states
One purpose of the V1000CT-2 thermalhydraulics benchmark was in general to test the capability of Computational fluid dynamics (CFD) codes to represent vessel thermal hydraulics and to analyze in particular the coolant mixing in the downcomer and lower plenum of the reactor vessel
The numerical grid model was generated with the grid generator ANSYS ICEMCFD and contains 4.7 Mio. tetrahedral elements

Summary

Introduction

Several mixing phenomena characterize the various operating conditions of pressurized water reactors (PWRs) and influence the safety analyses of the plant operating states. As a result phase 2 of the VVER-1000 coolant transient benchmark, [2] was defined aiming at mixing models testing and single effect analysis of main steam line break (MSLB) transients with improved vessel thermalhydraulic models. The experiment includes single-loop heating up by disturbing the heat transfer in the steam generator (SG) through the steam valves, at low reactor power in the range of 5–14% and with all main coolant pumps (MCPs) in operation. It was conducted during the plant commissioning phase at Kozloduy-6

The VVER-1000 Reactor Design

CFD Code and Sensitivity Analysis According to BPG

Computational Results

Conclusions