This study investigates the impact resistance of reinforced concrete structures under extreme loading conditions, with a particular focus on scenarios relevant to the safety of nuclear power plants. Since airplane crashes have been a critical design consideration since 1977, understanding the structural response of reinforced concrete under high-velocity impact is essential for ensuring safety and resilience. Large-scale impact tests are limited in availability and have been conducted infrequently, but they provide valuable insights that complement small-scale laboratory experiments, allowing researchers to validate scaling laws and improve predictive models. This research, conducted as part of a collaboration between Technische Universität Dresden (TUD/IMB) and the Bundesanstalt für Materialforschung und -prüfung (BAM), aims to bridge the gap between laboratory-scale experiments and real-world structural performance.The experimental setup involved reinforced concrete plates (3.0 m × 3.0 m × 0.4 m) tested in a four-point supported test setup under hard impact loading using a gravity-driven drop tower. A cylindrical steel impactor (404 kg) was released from a height of 9.5 m, achieving an impact velocity of about 13.6 m/s. The study employed high-precision measurement techniques, including load cells, accelerometers, triangulation lasers, and high-speed cameras with digital image correlation (DIC), to capture force, deformation, and failure characteristics.The results indicate a distinct progression of force transmission and deformation, with localized damage characterized by global deformation, a breakout cone and scabbing at the bottom surface of the plate. 3D laser scanning is used to document the outer damage pattern and create a digital image of the specimen after impact event. The experimental data confirmed a significant transfer of impact energy, leading to plastic deformations and possible reinforcement yielding. The findings contribute to a growing database for validating numerical models and refining impact-resistant design strategies for critical infrastructure, such as nuclear power plants. The test results will be further consolidated with additional tests in subsequent publications.
In a sodium-cooled fast reactor (SFR), thermal stratification, a sodium jet at the core outlet, and core–plenum interactions simultaneously occur in the reactor vessel (RV) under natural circulation (NC) decay heat removal when using a dipped-type direct heat exchanger (D-DHX). These localized three-dimensional (3D) thermal–hydraulic phenomena affect the temperature field in the RV, which consequently affects the NC flow rate in the primary loop. In turn, the NC flow feeds back into the temperature field in the RV. To accurately evaluate these mutual effects in the transient conditions of an SFR plant, a coupled analysis method combining the plant dynamics analysis code (system code) with the computational fluid dynamics (CFD) code for the 3D thermal hydraulics in the RV is necessary. In this study, we develop a coupled analysis framework by combining the Super-COPD system code for the primary loop flow with the CFD code for the RV using a coupling interface. In particular, we employ a CFD model for the RV (RV-CFD), which enables the simulation of 3D local phenomena in the RV with reduced computational cost while maintaining the prediction accuracy by incorporating the subchannel analysis model for assemblies. As a first step toward investigating the applicability of the coupled analysis method using Super-COPD and RV-CFD, we simulate a reactor scram transient behavior (from the rated operation to NC conditions) using the primary reactor auxiliary cooling system (PRACS) in the PLANDTL-1 sodium experimental apparatus. The result indicated that the coupled analysis successfully reproduced key RV thermal–hydraulic features, including thermal stratification and jet behavior, via feedback with the thermal hydraulics in the primary loop. In addition, the coupled analysis generally reproduces the temperature and flow behavior in the primary loop in agreement with experimental results. A comparison with overall plant analysis using Super-COPD alone reveals that the difference between the two approaches remains limited under the present PRACS operation case. This is attributed to the relatively small impact of 3D thermal hydraulics under this condition. However, the present study serves as an important first step toward applying the coupled analysis method to DRACS operation using D-DHX, where more complex phenomena are expected to occur in the RV.
Spacers in a BWR fuel assembly have an effect of improving the critical heat flux (CHF) by enhancing droplet deposition on the fuel rod surface. In this study, the behavior of droplets in the downstream region for a ferrule-type spacer and a vaned grid-type spacer was observed with a test section simulating the center subchannel of a BWR fuel assembly. The effects of the spacer on droplets were defined as the “turbulent promotion effect” and the “run-off effect”. The former refers to an effect on droplets in the dispersed flow due to turbulence downstream caused by the spacer. The latter refers to an effect on droplets generated from the liquid film on the spacer structure. The turbulent promotion effect affected droplets in a diameter standardized by the hydraulic equivalent diameter of the sub-channel less than 0.008 (the ferrule-type spacer) and 0.02 (the vaned grid-type spacer), and enhanced their deposition on the fuel rod surface downstream of the spacer. The droplet deposition distribution downstream of the spacer had two peaks, one in the immediate vicinity of the spacer and the other farther downstream (over 10 times the hydraulic equivalent diameter of the sub-channel), and both frequencies depended on the geometry. The run-off effect affected droplets generated by the spacer varying their outflow angle widely, and the droplet deposition distribution had a steep peak in the immediate vicinity of the spacer. Based on the experimental results, a mathematical model of droplet deposition enhancement was developed for each spacer type.
The Steam Explosion Modelling and Risk Analysis (SEMRA) code is developed to support the analysis of melt-coolant interactions and steam-explosion risk in water cooled nuclear reactors. In this work, the SEMRA shock-wave solver is extended to a variable-area Baer-Nunziato-type two-fluid model. The formulation incorporates velocity, thermal, and pressure relaxation processes, liquid-vapor mass transfer, buoyancy forces, and a Noble-Abel Stiffened-Gas (NASG) equation of state. The objective is to establish a computationally efficient and modular hydrodynamic solver for steam-explosion analysis that combines state-of-the-art numerical methods with flexible physical modelling. The aim of this work is to verify and validate the implemented physical and numerical capabilities and establish the hydrodynamic foundation for steam-explosion simulations.The extended SEMRA solver is assessed using analytical solutions, numerical benchmarks, and experimental data covering shock-wave propagation, low-Mach variable-area flow, cavitation, phase change, and buoyancy-driven motion. Two coupled flux formulations were evaluated: (i) Godunov and AUSM+ and (ii) Central-Upwind and Rusanov. The effect of thermodynamic modelling, mesh resolution, and geometric assumptions were also investigated. The results demonstrate that the developed formulation captures key multiphase flow behaviour and provides a robust framework for steam-explosion modelling applications.
