Volume 89, Issue 2 - Special Issue: The 34th German CFD Network of Competence Meeting

The 34th German CFD Network of Competence Meeting was held in March 2023 at the Gesellschaft für Anlagen-und Reaktorsicherheit (GRS) gGmbH in Garching, Germany. Twenty-one scientific presentations, distributed in the two sessions “Simulation of Reactor Cooling Circuit Flows” and “Simulation of Reactor Containment Flows” were presented. This paper gives a short summary of the different contributions.

The interaction between cooling fluid and solid structures (rods, tubes) in nuclear power plants may lead to flow-induced vibrations (FIV), causing material fatigue, fretting wear, and eventually loss of component integrity. This can cause further safety issues as well as substantial standstill costs due to longer or unplanned outages. With the growing computational power, the application of modern 3D numerical simulation tools for the accurate prediction of FIV phenomena is rapidly increasing. In 2022, the GO-VIKING ( G athering expertise O n V ibration I mpa K t I n N uclear power G eneration) project received a grant within the Horizon Europe research and innovation funding program. Sixteen European and two US partners started their collaboration in the field of FIV experiments and analysis. Over four years, the GO-VIKING project investigates FIV phenomena occurring in nuclear reactor cores and steam generators under single- and two-phase flow conditions. The project’s main objectives are to expand the expertise in the field of FIV through generation of new experimental and high-resolution numerical data; development, improvement, and validation of fluid-structure interaction (FSI) methods for FIV evaluation; training stakeholders in the application of these methods; and synthesizing guidelines for the prediction and assessment of FIV phenomena in nuclear reactors. This paper provides an overview of the GO-VIKING objectives, scientific program, as well as of the main scientific achievements in the first project year.

The investigation on heat transfer in post-dryout region is of great significance to determine the maximum wall temperature when dryout occurs. In this paper, the superheated vapor is considered as Eulerian continuous phase. With DPM (Discrete Particle Method) in the ASNSYS Fluent, droplets will be tracked with Lagrangian method. Heat, momentum and mass are exchanged between the two phases inside Eulerian control volumes. The stochastic tracking is included to investigate the effect of turbulence in the continuous phase on the droplet motion. The results show that the wall temperature profile differs a lot under different initial droplet sizes. By summary of the droplet evaporation rate, it’s found that less than 2 % evaporation happens directly on the wall surface, while evaporation mostly happens in the vapor layer near the wall.

Even though very unlikely to occur, severe accident scenarios in nuclear power plants have to be analyzed. During high-pressure core meltdown scenarios in a pressurized water reactor the primary circuit should fail first. Previous analyses found that a free convection flow within the vertical steam generator (SG) tubes with a simultaneous stratified gas counterflow in the hot legs could arise. This phenomenon leads to higher thermal loads on individual SG tubes which might then fail leading to a containment bypass and the release of radioactive material into the environment. Lumped parameter system codes used for safety analyses do not provide the models necessary to simulate phenomena like mixing in three-dimensional flows and could not consider local turbulence effects. Computational fluid dynamic (CFD) codes provide such capabilities but are much more computationally expensive. Coupling of a system code with a CFD code can therefore be used to simulate such phenomena. The advantages of both approaches can be maximized by splitting up the simulation domain between the codes, depending on the expected flow conditions. The system code AC 2 coupled with the CFD code OpenFOAM was used to simulate part of the severe accident transient. Free convection in the hot leg and the U-tubes of the vertical SG was observed in case of high-pressure severe accident sequences. The thermal load of individual SG tubes has been estimated from the results. These loads can be used as inputs for structural-mechanical analyses to estimate which part of the primary circuit would fail first.

System thermal-hydraulics (STH) codes are successfully used in the last decades for design and analysis of the physical behavior of nuclear power plants (NPPs). These programs provide reliable results at comparatively low computational cost, but have limited capabilities to predict complex 3D flow phenomena, which is possible with Computational Fluid Dynamics (CFD) tools. In order to combine the advantages of STH and CFD codes, algorithms to couple these are developed. At GRS, a new OpenFOAM solver, based on the volume of fluid (VOF) method, was developed together with new coupling boundary conditions to couple the CFD program OpenFOAM with the system code ATHLET. In a next step, verification and validation activities were carried out. These included the analysis of a Pressurized Thermal Shock (PTS) scenario. For this purpose, data from the OECD/NEA ROSA V Test 1.1, carried out in the Japanese Large-Scale Test Facility (LSTF) was used. The experiment investigated the temperature stratification after emergency core cooling (ECC) injection under single- and two-phase natural circulation conditions in the primary cooling circuit of a pressurized water reactor (PWR). This paper discusses the results of the coupled simulations, compares the numerical results with experimental data, and identifies areas for future improvements.

Because of its relevance for the safety analysis of pressurized water reactors (PWR), many research activities on flashing flows in pipes and nozzles arose from the mid of last century. Most of them have been focused on the critical mass flow rate and transient pressure or temperature fluctuations by means of experiments and system codes. Since the beginning of this century, owing to the increase in computer speed and capacity, computational fluid dynamics (CFD) is being used more and more in the investigation of flashing flows, which has the advantage of providing three-dimensional insights in the internal flow structure as well as its evolution. This work presents an overview of relevant flashing scenarios in the nuclear safety analysis, and focuses on the discussion about possibilities and challenges of using CFD modelling. It is shown that a two-fluid model with the thermal phase-change model is superior to a mixture model with pressure phase-change, relaxation and equilibrium models, respectively, in terms of interfacial mass transfer, however, efforts are still required to improve the interphase heat-transfer model. Furthermore, since flashing is accompanied with high void fraction and broad bubble size ranges, a poly-disperse two-fluid model is recommended, but the effect of phase change on bubble coalescence and breakup needs further research. In addition, during flashing the flow pattern may change from single phase to bubbly flow, churn flow, annular flow, and even mist flow. The rapid change of interfacial topology as well as its influence on the applicability of closure models remains a challenge.

The design of new nuclear reactor types as well as the analysis of certain phenomena in existing reactors require to consider different physical models and three-dimensional phenomena like the effect of turbulence in fluids and three-dimensional (3D) heat conduction in complex structures. One-dimensional (1D) lumped parameter system codes used to simulate transients in nuclear reactors lack high-resolution models. Special codes like computational fluid dynamics (CFD) or computational structural mechanics codes (CSM) can simulate those, but they require significantly more computational resources. Therefore, different coupling methods have been developed to limit the use of computationally expensive codes to those parts of a plant where they are really needed and couple them with systems codes, which simulate the rest. The library preCICE enables simultaneous coupling of multiple simulation programs, e.g., the 3D CFD code OpenFOAM and the 3D CSM code CalculiX. preCICE coupling interfaces were developed for the system code ATHLET. We focused on fluid-fluid couplings and conjugate heat-transfer couplings. Coupled simulations were performed for an experiment at the test facility TALL for the transition from forced to natural circulation coupling ATHLET with OpenFOAM, and a generic building condenser geometry coupling the three codes ATHLET, CalculiX, and OpenFOAM. The correct integral transfer of the quantities at the coupling interfaces was verified. However, it was found that coupled quantities at the interfaces need to be converted from 0D to 2D and backwards considering the underlying physics, which goes beyond the pure numerical considerations of the preCICE library.

CFD modelling of the thermo-hydraulic phenomena in the containment during the various phases of a severe accident necessarily requires consideration of radiative heat transfer – in the presence of steam. These radiative phenomena include (i) energy transfer within the gas mixture and (ii) between the gas and surrounding structures. Preliminary calculations carried out for these types of experiments within the OECD/NEA HYMERES-2 project with the CFD code containmentFOAM using a Monte Carlo solver for thermal radiation, demonstrated that the radiative heat transfer is significant even for very small amounts of vapour in the range of ≈0.1 % to ≈2 %. For this reason, the test matrix was tailored to the two opposite extremes: either gas compositions with a low vapour/steam content, where radiative heat transfer can be neglected, or gas mixtures with higher vapour contents, so that radiative heat transfer plays a dominant role. For the selected experiments of the H2P2 series and the corresponding CFD calculations, a vessel with a diameter of 4 m and a height of 8 m was preconditioned with different air-vapour mixtures (a) at room temperature and (b) elevated temperatures. A stable helium layer was then built-up in the upper part of the vessel. The gas was then compressed by injecting helium from above which resembles with best efforts a compression with a piston in a cylinder. This results in a height-dependent and transient increase of the gas temperature. These experiments and the associated CFD calculations were developed to isolate the phenomena of thermal radiation as good as possible from convective and diffusive effects – within the always present experimental limitations. For the reference experiment with ‘dry conditions’ corresponding to the lowest experimentally possible humidity of ≈0.1 %, we show that the use of a model without radiation provides the best agreement between the experimental and numerical results. For the much higher steam content of ≈60 %, the statistical narrow band correlated-k model (SNBCK), non-gray gas model, is the best candidate for future calculations – with computationally forgivable additional effort. We also provide with the Filtered Rayleigh Scattering technique (FRS) an outlook for a possible future instrumentation approach to better meet the requirements of the CFD community.

In the case of a severe accident in a nuclear power plant, hydrogen could be produced in the superheated reactor core. The hydrogen could flow through a leak in the primary circuit into the containment and a hydrogen deflagration could occur. A consequence of such an event was the Fukushima accident in 2011. For the analysis of such events, different phenomena in the reactor containment must be simulated correctly such as the distribution of steam and hydrogen, steam condensation, thermal radiation, conjugated heat transfer, hydrogen recombination and deflagration. The application of Computational Fluid Dynamics (CFD) methods for such analyses is becoming more relevant. Selected models needed to simulate the different containment phenomena with OpenFOAM are currently being tested, validated or improved and further developed by a network of German research institutes and universities.