Soeren Kliem

The development, verification, and validation of Computational Fluid Dynamics (CFD) codes in reference to nuclear power plant (NPP) safety has been a focus of many research organizations over the last few decades. Therefore, a collection of Rossendorf Coolant Mixing Test Facility (ROCOM) CFD-grade experiments was made obtainable to line up a global International Atomic Energy Agency (IAEA) benchmark regarding Pressurized Thermal Shock (PTS) situations. The benchmark experiment describes the complicated flow structures in mixed convection zones of the RPV during PTS events. The experiments were utilized to validate CFD codes. Additionally, an experiment with no buoyancy forces was elite to point out the influence of density variations. Compared to earlier studies, the turbulence models of the CFD code improved a lot. The turbulence modeling approach shows a respectable agreement with the experimental data.

This paper describes the main objectives, technical content, and status of the H2020 project entitled “High-performance advanced methods and experimental investigations for the safety evaluation of generic Small Modular Reactors (McSAFER)”. The main pillars of this project are the combination of safety-relevant thermal hydraulic experiments and numerical simulations of different approaches for safety evaluations of light water-cooled Small Modular Reactors (SMR). It describes the goals, the consortium, and the involved thermal hydraulic test facilities, e.g., the COSMOS-H (KIT), HWAT (KTH), and MOTEL (LUT), including the experimental programs. It also outlines the different safety assessment methodologies applied to four different SMR-designs, namely the CAREM (CNEA), SMART (KAERI), F-SMR (CEA), and NuScale. These methodologies are multiscale thermal hydraulics, conventional, low order, and high fidelity neutron physical methods used to demonstrate the inherent safety features of SMR-core designs under postulated design-basis-accident conditions. Finally, the status of the investigations is shortly discussed followed by the dissemination activities and an outlook.

The aim of the numerical study was the detection of possible vortices in the upper part of the core of a Pre-Konvoi Pressurized Water Reactor (PWR) which could lead to temperature cycling. In addition, the practical application of this Computational Fluid Dynamic (CFD) simulation exists in the full 3D analysis of the coolant flow behavior in the reactor pressure vessel of a nuclear PWR. It also helps to improve the design of future reactor types. Therefore, a CFD simulation of the flow conditions was carried out based on a complex 3D model. The geometry of the model includes the entire Reactor Pressure Vessel (RPV) plus all relevant internals. The core is modelled using the porous body approach, the different pressure losses along and transverse to the main flow direction were considered. The spacer-grid levels were taken into account to the extent that in these areas no cross-flow is possible. The calculation was carried out for nominal operating conditions, i.e., for full load operation. Furthermore, a prototypical End of Cycle (EOC) power distribution was assumed. For this, a power distribution was applied as obtained from a stationary full-core calculation with the 3D neutron kinetics code DYN3D. In order to be able to adequately reproduce flow vortexes, the calculation was performed transiently with suitable Detached Eddy Simulations (DES) turbulence models. The calculation showed fluctuating transverse flow in the upper part of the core, starting at the 8th spacer grid but also revealed that no large dominant vortices exists in this region. It seems that the core acts as a rectifier attenuating large-scale vortices. The analyses included several spacer grid levels in the core and showed that in some areas of the core cross-section an upward increasingly directed transversal flow to the outlet nozzle occurs. In other areas of the core cross-section, on the other hand, there is nearly any cross-flow. However, the following limitations of the model apply: In the model all fuel elements are treated identical and cross flows due to different axial pressure losses for different FA types cannot be displayed. The complex structure of the FAs (eg. flow vanes in spacer grids) could also influence the formation of large-scale vortices. Also, the possible influence of two-phase flows was not considered.