Contact

Prof. Dr.-Ing. Dr. h. c. Uwe Hampel

Director Institute of Fluid Dynamics
u.hampel@hzdr.de
Phone: +49 351 260 2772

Dr. Sebastian Unger

Head Thermal Energy and Process Technology
s.ungerhzdr.de
Phone: +49 351 260 3225

Peter Schütz

p.schuetzhzdr.de
Phone: +49 351 260 3286
+49 351 260 2865

Thermohydraulic Nuclear Safety Research

Future nuclear power plants will be equipped with advanced passive safety systems. Such systems — for example, for cooling processes — do not require external power or signal processing and operate independently of human intervention or moving components such as pumps. Instead, the driving forces are based on fundamental physical principles. For instance, convective heat transfer can be achieved solely through natural circulation driven by gravity, density differences, or pressure gradients.

Experimental setup for pressure equalization to investigate the fluid dynamics of thermal shock in emergency cooling scenarios.

In this context, multiphase mixtures (e.g., steam/water flows) occur, which are known for their complex mass, heat, and momentum transfer mechanisms. Advanced system codes or CFD (Computational Fluid Dynamics) codes can only accurately predict such phenomena if the underlying models have been validated with experimental data. Therefore, the experimental characterization of multiphase phenomena must be carried out under operating conditions and scales that are as close as possible to real applications, while also capturing a wide range of flow regimes.

For this purpose, application-oriented experiments are conducted at the TOPFLOW facility at pressures of up to 70 bar and temperatures of up to 286 °C. The goal of these experimental investigations is to capture the relevant phenomena and processes, thereby generating a database for the validation of numerical codes. These validated codes are then used to predict flow behavior in nuclear power plants and to support the safe design of future plant processes.

TOPFLOW

Complex flow phenomena are investigated here under realistic conditions to enable energy efficient material flows in process engineering and the chemical industry as well as the safe operation of nuclear power plants.

Gas liquid flows in vertical pipes

In the two vertical test sections of the TOPFLOW experimental facility (DN50, DN200), flow regimes in adiabatic (water/air) and non-adiabatic (water/water vapor) flows are investigated. With the aid of imaging measurement techniques, validation data can thus be derived and individual mechanisms can be examined.

Two-phase flow around obstacles

Experiments on bubbly flows in a constricted pipe, in which detailed information on phase distributions, bubble properties, and liquid velocities was obtained using ultra-high-resolution X-ray tomography and hot-wire anemometry for the further development and validation of numerical models.

Counter-current flow limitation during reflux condenser in the hot leg of a PWR nuclear reactor

During emergency cooling scenarios, countercurrent flow of steam and condensate can occur in the hot leg of pressurized water reactors. To simulate instabilities of such flows, fundamental experiments on countercurrent flow limitation were conducted at HZDR.

Thermal fluid dynamics of pressurized thermal shock

When cold emergency cooling water is injected into the main coolant line of a nuclear reactor, locally high thermomechanical stresses occur due to temperature differences. In an experimental study, heat and phase-change transfer processes were analyzed and provided as numerical validation data.

Foto: Simplified Representation of the Experimental Basin in the Pressure Vessel of the TOPFLOW Facility ©Copyright: Dr. Alexander Döß

Fluid dynamics of direct steam condensation under pressure (DENISE)

Condensation effects occur at the interface between cold cooling water and saturated steam. The governing mechanisms were clarified in single-effect studies on stratified flows, free jets, and gas entrainment. For this purpose, imaging measurement techniques were qualified at operating pressures of up to 50 bar.

Single-effect studies on steam condensation in an inclined pipe

A drop in the water level in the reactor pressure vessel caused by accident scenarios is to be counteracted by a passive emergency condensation system. The flow regimes and phase distributions occurring in the slightly inclined condensation pipes were investigated using an imaging tomography system, and heat transfer was measured with a specially developed probe.