Nuclear Reactor Safety
The work aims at the development and application of methods for the analyses of transients and postulated accidents in currently operating nuclear reactors and in new reactor types being under development in the neighboring countries. Emphasis is put on the reactor dynamics code DYN3D, which has been developed for the analyses of design basis accidents in light water reactors. DYN3D is applicable to reactors with quadratic and hexagonal fuel assembly cross section. It has been coupled with advanced system codes to broaden the spectrum of transients and accidents that can be analyzed. 13 organizations in Germany and Europe are users of the DYN3D code, among them nuclear safety authorities, research and technical support organizations and universities. DYN3D was implemented into the European code-platform NURESIM and serves as one of the reference codes for core calculations in Europe. A transport option (SP3) for neutron flux calculation was implemented for quadratic fuel assemblies in order to improve the accuracy of the code results.
DYN3D is also available for Molten Salt Nuclear Reactors. A version for the gas-cooled high temperature reactor is nearly finished. The main improvement concerns a 3D heat transport model of the prismatic graphite fuel blocks.
Reactor physics studies are carried out with the goal to obtain core designs with fuels dedicated to the reduction of actinide production. This work is devoted to existing and the above mentioned future reactors.
Multi-phase flows occur in many industrial processes, but are of special importance in safety analyses of nuclear reactors. The available simulation tools still have considerable deficiencies to predict transient multi-phase flows. The reliability of prediction of the spatial and temporal flow characteristics does not yet meet the requirements for nuclear safety analyses.
Accordingly, the work within this theme aims at the qualification of Computational Fluid Dynamics (CFD) codes for two-phase flows under conditions relevant to nuclear safety. The corresponding activities on the development and validation of local, physics-based, geometry-independent closure models reflecting the mass-, momentum and energy transfer between the phases are concentrated within the Division of Computational Fluid Dynamics of the Institute of Fluiddynamics. The single research topics follow the requirements defined by the German CFD-network in Nuclear Safety Research. The theoretical work is based on experimental data, which have to be of high resolution in space and time. For this reason innovative measuring techniques, including tomographic methods have to be adopted. Such experiments and the development of new measuring techniques are conducted in the Division of Experimental Thermal Fluid Dynamics. Recently, data were obtained using the wire-mesh sensor technology e.g. on bubbly flows with phase transfer in a large vertical pipe or on counter-current flow limitation in a model of the hot leg of a PWR. These data are now used for CFD model qualification as e.g. for the extension of the previously developed Multi-Bubble-Size-Group (MUSIG) model to situations involving phase transfer. In future such data will also be obtained by non-intrusive ultrafast X-ray tomography.
CFD models are also created and used for the analyses of boron dilution accidents and accidents with the transport of mineral wool into the reactor pressure vessel. The coolant mixing in the primary circuit of PWRs is experimentally investigated at the ROCOM test facility.
Reactor construction materials undergo degradation due to fast neutron irradiation. It is known for the reactor pressure vessel (RPV) materials of light water reactors that nano-scaled defects primarily induced by collision cascades caused by the fast neutrons result in a decrease of ductility. The steel gets brittle. During loss of coolant accidents, cold water injection into the downcomer may provoke considerable thermal stresses in the RPV wall which have to be relieved by ductile strain to withstand the load and to keep the integrity of the RPV. The fundamental ageing phenomena connected with fast neutron irradiation considering for example the flux effect as well as the materials composition have to be investigated to render possible the consistent multi-scale modelling of the materials behaviour under high fast neutron loads. The work on these topics is conducted in the Division of Structural Materials of the Institute of Ion Beam Physics and Materials Research. In international collaboration (EU projects: LONGLIFE, PERFORM60) the multi-scale modelling chain has to be closed and harmonised codes and standards for ageing monitoring have to be established. Further, the neutron irradiation induced ageing of reactor pressure vessel materials is explored. An extensive research project is being devoted to the post-irradiation examination of the reactor pressure vessels of the decommissioned Greifswald WWER-440 reactors.
- Development of Multi-Physics Code Systems based on the Reactor Dynamics Code DYN3D;
S. Kliem, A. Gommlich, A. Grahn, U. Rohde, J. Schütze, T. Frank, A. Gomez, V. Sanchez, Kerntechnik 76(2011)3, KT100569
- Experiments on slug mixing under natural circulation conditions at the ROCOM test facility using high resolution measurement technique and numerical modeling;
S. Kliem, T. Höhne, U. Rohde, F.-P. Weiss, Nuclear Engineering and Design 240(2010)9, 2271-2280