Evaluation of CFD Capability for Simulation of Energetic Flow in Light Water Reactor Containment (original) (raw)

Evaluation of computational fluid dynamic methods for reactor safety analysis (ECORA)

Nuclear Engineering and Design, 2005

This report addresses the activities in the field of CFD (Computational Fluid Dynamics) to simulate processes, which can occur in a nuclear containment. The report starts with a summary of the phenomena to be relevant under abnormal and accidental conditions. A classification of the known issues is intended to help in the evaluation of the current level of modelling such issues and may give some indication for a continuation of work.

Application of CFD Codes in Nuclear Reactor Safety Analysis

… and Technology of …, 2010

Computational Fluid Dynamics (CFD) is increasingly being used in nuclear reactor safety (NRS) analyses as a tool that enables safety relevant phenomena occurring in the reactor coolant system to be described in more detail. Numerical investigations on single phase coolant mixing in Pressurised Water Reactors (PWR) have been performed at the FZD for almost a decade. The work is aimed at describing the mixing phenomena relevant for both safety analysis, particularly in steam line break and boron dilution scenarios, and mixing phenomena of interest for economical operation and the structural integrity. For the experimental investigation of horizontal two phase flows, different non pressurized channels and the TOPFLOW Hot Leg model in a pressure chamber was build and simulated with ANSYS CFX. In a common project between the University of Applied Sciences Zittau/Görlitz and FZD the behaviour of insulation material released by a LOCA released into the containment and might compromise the long term emergency cooling systems is investigated. Moreover, the actual capability of CFD is shown to contribute to fuel rod bundle design with a good CHF performance.

Direct Containment Heating Investigations for European Pressurized Water Reactors

While the issue on Direct Containment Heating (DCH) was resolved for US reactor plants in the 1990s it was found that the consequences of DCH processes are strongly dependent on the reactor cavity configuration. Therefore, an experimental and analytical program was started in 1998 to inves-tigate melt ejection scenarios for typical German and European reactor designs. Six experiments have been performed in a 1:18 scaled reactor geometry, characterized by a narrow pit without exit other than through the annular space between pressure vessel and cavity wall, leading either directly to the upper containment or into the pump and steam generator rooms along the flow path around the main cooling lines. The corium was modeled by an iron-alumina melt that was driven by steam, and a pro-totypic atmosphere in the containment was applied. Since the system pressure at core melt accidents will be low due to compulsory system depressurization, the vessel failure pressures were kept between 0.8 an...

The Use of CFD Code for Numerical Simulation Study on the Air/Water Countercurrent Flow Limitation in Nuclear Reactors

2015

For the experimental study on the air/water countercurrent flow limitation in Nuclear Reactors, were built at CDTN an acrylic test sections with the same geometric shape of "hot leg" of a Pressurized Water Reactor (PWR). The hydraulic circuit is designed to be used with air and water at pressures near to atmospheric and ambient temperature. Due to the complexity of the CCFL experimental, the numerical simulation has been used. The aim of the numerical simulations is the validation of experimental data. It is a global trend, the use of computational fluid dynamics (CFD) modeling and prediction of physical phenomena related to heat transfer in nuclear reactors. The most used CFD codes are: FLUENT®, STARCD®, Open Foam® and CFX®. In CFD, closure models are required that must be validated, especially if they are to be applied to nuclear reactor safety. The ThermalHydraulics Laboratory of CDTN offers computing infrastructure and license to use commercial code CFX®. This article ...

CFD model of fluid flow in reactor : A simulation of velocity and heat distribution in a channel

2006

The basic problem of operating a boiling water nuclear reactor (BWR) is that of maximizing the power output while avoiding fuel rod over-heating (dry-out). For the safe operation of BWRs this entails a detailed understanding of the flow of water and steam through the reactor core. In a BWR the water functions not only as a coolant, but also as a moderator for the neutrons emitted in the fission process. To describe the thermohydraulic properties of the reactor a number of parameters are of common interest. Examples of such parameters are void, pressure, temperature, water and steam velocities, pressure, sheer forces and turbulent kinetic energy. There are a few ways of revealing these values such as experiments built up to behave like a reactors and computer simulations using models based on the laws of fluid dynamics and thermodynamics. This research concerns a computer based model which uses a continuous fluid dynamic, (CFD), calculation program called OpenFOAM (Open Field Operati...

Analysis of Ex-Vessel Steam Explosion with MC3D

An ex-vessel steam explosion may occur when, during a severe reactor accident, the reactor vessel fails and the molten core pours into the water in the reactor cavity. A steam explosion is a fuel coolant interaction process where the heat transfer from the melt to water is so intense and rapid that the timescale for heat transfer is shorter than the timescale for pressure relief. This can lead to the formation of shock waves and production of missiles that may endanger surrounding structures. A strong enough steam explosion in a nuclear power plant could jeopardize the containment integrity and so lead to a direct release of radioactive material to the environment. In the paper, different scenarios of ex-vessel steam explosions in a typical pressurized water reactor cavity are analyzed with the code MC3D, which was developed for the simulation of fuel-coolant interactions. A comprehensive parametric study was performed varying the location of the melt release (central, left and right side melt pour), the cavity water subcooling, the primary system overpressure at vessel failure and the triggering time for explosion calculations. The main purpose of the study was to determine the most challenging ex-vessel steam explosion cases in a typical pressurized water reactor and to estimate the expected pressure loadings on the cavity walls. The performed analysis shows that for some ex-vessel steam explosion scenarios significantly higher pressure loads are predicted than obtained in the OECD programme SERENA Phase 1.

An approach to numerical simulation and analysis of molten corium coolability in a boiling water reactor lower head

2010

This paper discusses an approach for application of the computational fluid dynamics (CFD) method to support development and validation of computationally effective methods for safety analysis, on the example of molten corium coolability in a BWR lower head. The approach consists of five steps designed to ensure physical soundness of the effective method simulation results: (i) analysis and decomposition of a severe accident problem into a set of separate-effect phenomena, (ii) validation of the CFD models on relevant separate-effect experiments for the reactor prototypical ranges of governing parameters, (iii) development of effective models and closures on the base of physical insights gained from relevant experiments and CFD simulations, (iv) using data from the integral experiments and CFD simulations performed under reactor prototypic conditions for validation of the effective model with quantification of uncertainty in the prediction results and (v) application of the computationally effective model to simulate and analyze the severe accident transient under consideration, including sensitivity and uncertainty analysis. Implementation of the approach is illustrated on a so-called effective convectivity model for simulation of turbulent natural convection heat transfer and phase changes in a decay-heated corium pool. It is shown that detailed information obtained from the CFD simulations are instrumental to ensure the effective models capture safety-significant local phenomena, e.g. the enhanced downward heat flux in the vicinity of a cooled control rod guide tube.

CFD Simulations for Safety of Chemical Reactors and Storage Tanks

Chemical Engineering & Technology, 2011

Two CFD models have been formulated for the stirred tank reactor and the storage tank, respectively. Spatial and time profiles of the liquid velocity, temperature and concentration inside the reactor and storage tank have been determined for different thermal runaway scenarios. The obtained results indicate that the CFD models can be very useful to elaborate an efficient and robust method for on-line early runaway detection.

Computational Fluid Dynamics Applied to Study Coolant Loss Regimes in Very High Temperature Reactors

Brazilian Journal of Radiation Sciences

The nuclear energy is a good alternative to meet the continuous increase in world energy demand. In this pers-pective, VHTRs (Very High Temperature Reactors) are serious candidates for energy generation due to its inherently safe performance, low power density and high conversion efficiency. However, the viability of these reactors depends on an efficient safety system in the operation of nuclear plants. The HTR (High Temperature Reactor)-10 model, an experimental reactor of the pebble bed type, is used as a case study in this work to perform the thermohydraulic simulation. Due to the complex patterns flow that appear in the pebble bed reactor core CFD (Computational Fluid Dynamics) techniques are used to simulate these reactors. A realistic approach is adopted to simulate the central annular column of the reactor core. As geometrical model of the fuel elements was selected the BCC (Body Centered Cubic) arrangement. Parameters considered for reactor design are available in the tec...