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.
- Pressure Vessels and Piping Division
Determination of Thermal-Hydraulic Loads on Reactor Internals in a DBA-Situation
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Lestinen, V, Toppila, T, Timperi, A, Pa¨ttikangas, T, & Ha¨nninen, M. "Determination of Thermal-Hydraulic Loads on Reactor Internals in a DBA-Situation." Proceedings of the ASME 2006 Pressure Vessels and Piping/ICPVT-11 Conference. Volume 4: Fluid Structure Interaction, Parts A and B. Vancouver, BC, Canada. July 23–27, 2006. pp. 1539-1548. ASME. https://doi.org/10.1115/PVP2006-ICPVT-11-93456
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