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In the hypothetical Large-Break Loss Of Coolant Accident (LBLOCA), rapid depressurization of the reactor primary circuit causes loads on the reactor internals. This paper presents numerical simulations of a HDR experiment, where LBLOCA of a pressurized water reactor due to a sudden pipe break in the primary loop was studied. In the experiment, Fluid-Structure Interaction (FSI) phenomena caused by the flexibility of the core barrel were studied in particular. Star-CD Computational Fluid Dynamics (CFD) code and ABAQUS structural analysis code were used for three-dimensional calculations. The MpCCI code was used for two-way coupling of the CFD and structural analysis codes in order to take FSI into account. Two-way FSI calculation was also performed with ABAQUS only by modeling water as an acoustic medium. Pressure boundary condition at the pipe break was evaluated with the system code APROS as a two-phase calculation. Comparisons with the experiment were made for fluid pressures and break mass flow as well as for structural displacements and strains. Fairly good agreement was found between the experiment and simulation when coupling of the CFD and structural analysis codes was used. For the acoustic calculation, the results showed good agreement in the early phase of the simulation. In the late phase, structural loads were over-predicted by the acoustic calculation due to the effect of bulk flow of water which is not included in the acoustic model.

According to Finnish regulatory requirements, reactor internals have to stay intact in design basis accident (DBA) situations, so that control rods can always penetrate into the core. This is the basic motivation to study and develop more detailed methods for analyses of thermal-hydraulic loads on reactor internals during the DBA situation in the Loviisa Nuclear Power Plant (NPP) in Finland. In this work, the studied accident situation is Large Break Loss of Coolant Accident (LBLOCA). The objective of this work is to connect thermal-hydraulic and mechanical analysis methods with the goal to produce a reliable method for determination of thermal-hydraulic and mechanical loads on reactor internals in the accident situation. In the present model, the downcomer of a PWR is only included and the reactor internals will be added later. The tools studied are thermal-hydraulic system codes, computational fluid dynamics (CFD) codes and finite element analysis (FEA) codes. Both thermal-hydraulic and mechanical aspects are discussed in this paper. Firstly, the pressure boundary condition in the pipe break point was calculated with the system code. In the second step, CFD analyses were made. Finally, the full fluid-structure interaction coupling between the CFD and FEA codes was used. The codes used for development and simulations are APROS system code for boundary condition calculations, STAR-CD and FLUENT for CFD calculations and ABAQUS for FEA calculations.

Recently, Small Modular Reactors (SMRs) have attracted increased public discussion. While large nuclear power plant new build projects are facing challenges, the focus of attention is turning to small modular reactors. This paper is based on the PhD Dissertation “Licensing Model Development for Small Modular Reactors (SMRs) - Focusing on Finnsh Regulatory Framework”, approved in 2013. The result of the study is a recommendation for a new, optimized licensing process for SMRs. The most important SMR-specific feature, in terms of licensing, is the modularity of the design. Here the modularity indicates multi-module SMR designs, which creates new challenges in the licensing process. The application of the new licensing process is developed using Systems Engineering, Requirements Management, and Project Management practices and tools.