APPLICATION OF COMPUTATION FLUID DYNAMICS IN NUCLEAR REACTOR SAFETY ANALYSIS
24th Symposium of AER on VVER Reactor Physics and Reactor Safety (2014, Sochi, Russia)
Nuclear applications of computational fluid dynamics (CFD)
APPLICATION OF COMPUTATIONAL FLUID DYNAMICS IN NUCLEAR REACTOR SAFETY ANALYSIS
Helmholtz-Zentrum Dresden – Rossendorf (HZDR) Bautzner Landstraße 400 | D-01328 Dresden Institute of Fluid Dynamics
The last decade has seen an increasing use of three-dimensional CFD codes to predict steady state and transient flows in nuclear reactors because a number of important phenomena such as pressurized thermal shocks, coolant mixing, and thermal striping cannot be predicted by traditional one-dimensional system codes with the required accuracy and spatial resolution. CFD codes contain models for simulating turbulence, heat transfer, multi-phase flows, and chemical reactions. Such models must be validated before they can be used with sufficient confidence in NRS applications. The necessary validation is performed by comparing model results against measured data. However, in order to obtain a reliable model assessment, CFD simulations for validation purposes must satisfy strict quality criteria given in the Best Practice Guidelines (BPG).
Numerical investigations on single phase coolant mixing in Pressurised Water Reactors (PWR) have been performed at the HZDR for almost a decade. The work is aimed at describing the mixing phenomena relevant for both safety analysis, particularly in steam line break and boron dilution scenarios, and mixing phenomena of interest for economical operation and the structural integrity.
On the other hand slug flow as a multiphase flow regime can occur in the cold legs of pressurized water reactors, for instance after a small break Loss of Coolant Accident (SB- LOCA). Slug flow is potentially hazardous to the structure of the system due to the strong oscillating pressure levels formed behind the liquid slugs. For the experimental investigation of horizontal two phase flows, different non pressurized channels and the TOPFLOW Hot Leg model in a pressure chamber was build and simulated with ANSYS CFX.
The behavior of insulation material released by a LOCA into the containment and the reactor core might compromise the long term emergency cooling systems. Subsequently, if the ECCS is operating in the sump recirculation mode, the debris suspended in the containment sump would begin to accumulate on the sump strainers. A small part could penetrate through the strainers and could be transported towards the reactor core. It was the aim of the numerical simulations to study where and how many mineral wool fibres are deposited at the upper spacer grid of a core.