Volume 78, Issue 1

Computational fluid dynamics simulations for two-phase flows are important in different fields of engineering and science. Since two-phase flows are inherently complex, also CFD modeling development requires special attention. The validation of model implementation and the derivation of physics based models for momentum, heat, and mass transfer in two-phase flow require experiments with generation of high-resolution measurement data. This, however, is a great challenge, since most standard flow measurement tools used in single phase flow situations, are not suited for multiphase flows. In this article we report on advanced imaging and measuring methods for two-phase flow experiments, which have been extensively used in the recent past to conduct experiments for two-phase flows at the Helmholtz-Zentrum Dresden-Rossendorf. In particular the application of wire-mesh sensors, ultrafast X-ray tomography, gamma ray tomography and positron emission tomography will be introduced and discussed.

Condensation-induced water hammers (CIWH) are consequences of rapid condensation of steam in partially filled pipes. They present a particular hazard potential, as they seem to occur spontaneously and out of stagnation. The entire process still eludes a secured analytical or numerical predictability and determination of the fluid dynamic loads. The simulation of these processes is one of the most difficult tasks of thermal hydraulic transient analyses. Condensation induced water hammers in plants – mostly power plants – can lead to significant costs if they result in long downtimes, detailed analyses and upgrades. In this contribution, the phenomenon CIWH is explained and the relevance for nuclear engineering will be discussed. An outlook on the actual requirements of regulatory guidelines and the state of science and technology will be given.

Condensation-induced water hammers belong to the most serious and complex pressure surges. Experimental data from literature are scanty, incomplete, not public or simply missing. A new test facility at the UniBw (University of the German Armed Forces) was constructed to perform own experiments within a research alliance. This new test facility uses a complete new approach. In contrast to other test stations, the UniBw water hammer test facility is a closed system. This leads on the one hand to a more complex handling with respect of conducting experiments but on the other hand also to a more realistic characterization of thermo-hydraulic phenomena which can occur in a power plant. The first results of a measured condensation-induced water hammer are presented.

A Hybrid Heat Transfer Coefficient module has been developed based on two Surface Renewal Theory models using CFD simulations. The validation of the model has been done on a meso-scale computational grid for CFD simulations and on a macro-scale computational grid for System Code analysis. The CFD simulation was performed for a stratified co-current two phase flow between saturated steam and sub-cooled water while the System Code analysis was performed for a Condensation Induced Water Hammer experiment.

Condensation induced water hammer (CIWH) are caused by contact of steam and subcooled water. Thus, modeling the direct contact condensation is a crucial step towards the simulation of condensation induced water hammer with 1D pressure surge codes. Therefore, also the TÜV NORD SysTec GmbH & Co. KG inhouse pressure surge code DYVRO has been equipped with a new contact condensation model. The validation of DYVRO against an experiment dealing with CIWH is presented in this contribution.

A collaborative project funded by the Federal Ministry of Education and Research (BMBF) in the framework of the R&D program “Energie 2020+” by four Universities, two Research Centres and ANSYS is coordinated by Helmholtz-Zentrum Dresden-Rossendorf (HZDR). The project is directed towards the development and validation of CFD models of boiling processes in PWR in the range from subcooled nucleate boiling up to the critical heat flux. The work is oriented towards basic research and focused on model development, simulation and experiments on different scales. An important objective is the education and maintaining of competence in nuclear technology.

The NUPEC “PWR Subchannel and Bundle Tests” (PSBT) were designed to investigate critical heat flux phenomena in rod bundles and yield data to validate hot channel codes. Some of the tests are suitable for validating a CFD wall boiling model. The paper presents CFD calculations of the void distribution tests of the PSBT benchmark using ANSYS CFX-12.1. First, relevant aspects of the implemented wall boiling model are reviewed highlighting the uncertainties in several model parameters. It is then shown that the measured cross sectionally averaged values can be reproduced well with a single set of calibrated model parameters for different tests cases. For the reproduction of patterns of void distribution cross sections attention has to be focussed on the modelling of turbulence in the narrow channel. Only a turbulence model with the capability to resolve turbulent secondary flows is able to reproduce at least qualitatively the observed void distribution patterns.

In this paper new developments in the framework of a common project simulating subcooled boiling are described. The computational model used combines the Euler/Euler two-phase flow description with heat flux partitioning. Main achievements were a comprehensive study of the boiling process itself and a better description of the interfacial area by coupling of wall boiling with a population balance model. The model extensions are validated and the present capabilities of CFD for wall boiling are investigated.

The paper presents results on the dynamics of bubbles in a mono- and in a bi-disperse bubble swarm. The radial pair correlation function provides insight into the characteristic bubble clustering length. The angular pair correlation function shows that for small distances the most probable alignment of a bubble pair is horizontal.

The paper presents a Large Eddy Simulation at high Reynolds number of a section of a sub-channel in a fuel rod-bundle assembly of square pattern inside a pressurized water reactor. The focus is on the instability generated by the geometry of the sub-channel.

To investigate boiling processes, a test facility has been constructed, which allows the use of optical techniques for the detection of individual vapor bubbles. Demineralized water was used as working fluid at ambient pressure whereat temperature, flow rate and heat flux density can be varied. The growth of individual vapor bubbles was recorded by high-speed camera and analyzed by digital image processing (DIP). Optical coherence tomography (OCT) enabled the visualization of steam bubbles in cross-sections perpendicular to the heated surface.

Critical heat flux (CHF) is an important effect in boiling heat transfer, but the effects causing it still remain to be fully understood. An overview of current mechanistic CHF models is presented, and experiments for validation are discussed.

An experimental facility for the study of boiling flows in a 3 × 3 rod bundle geometry was setup. The bundle resembles in essential geometrical parts the geometry in a pressurized water reactor fuel element. The facility is operated with a refrigerant fluid. Beside standard instrumentation for temperature, pressure and flow rate we employed particle image velocimetry for single phase flow studies, gamma ray densitometry for integral gas fraction measurement sand ultrafast X-ray tomography for the study of the void dynamics in the cross-section. Moreover extensive thermo-instrumentation allows axial rod surface temperature measurements for the central heated rod. First results will be discussed in this article.

Turbulent single-phase simulations in a geometrical domain reproducing the new ROFEX test facility (3×3 rod bundle) were conducted. These investigations were aimed to perform an initial study for future analysis of boiling processes in the same test facility. The final goal of these experiments in ROFEX is the validation of CFD boiling models. The current CFD simulations allow an assessment of the turbulent flow conditions in this test facility.

The present study investigates the effect of boric acid and lithium hydroxide on flow boiling phenomena in pressurized water reactors. The experiments were conducted in a boiling loop with up to 40 bar and 250 °C. The bubble size distribution was determined with a wire-mesh sensor. The results showed no indication that reactor-typical concentrations of boric acid or lithium hydroxide have a safety-relevant influence on the boiling behavior in pressurized water reactors.

In the course of a severe accident in a nuclear power plant the integrity of the reactor vessel may be preserved by External Reactor Vessel Cooling (ERVC). By flooding the cavity surrounding the reactor vessel the decay heat is removed from molten core material through the vessel wall to water at the submerged vessel surface. In order to simulate the boiling process at the external vessel wall during ERVC in the framework of lumped parameter codes a model based on a boiling model used in the code ANSYS CFX 13.0 is developed.