fluid flow behaviour during free vibrations of a mono-hexagonassembly validation of 3-d navier-stokes model in cast3m withexperiments on pise-1a

International audienceIn the scheme of French ASTRID (Advanced Sodium TechnologicalReactor for Industrial Demonstration) project, fluidstructure interaction phenomenon involved in the dynamic behaviourof core flowering, which could happen during seismicevents, are of high interest. Also core flowering behaviour isconsidered as the main initiating event for the four SCRAMsthat happened in PHENIX reactor during 1989 to 1990. Monoassemblytesting facility PISE-1A and multi-assembly experimentalmock-up PISE-2C were built in CEA for further researchespecially focusing on damping. A series of free-vibration experimentswith different conditions has been performed on PISE-1A.Corresponding 3-D simulations with Navier-Stokes model implementedwith CAST3M code have been performed for interpretation,validation and verification with the experiments. Addedmass effects on damping and frequency involved in the behaviourof the assemblys motion during free vibration tests with differentphysical conditions, including liquid height, viscosity and density,have been studied. Uncertainty sources from non-linearitiesduring experiments, data analysis methodologies and simulationstability have been investigated

Numerical simulations of a transient injection flow at low Mach number regime

International audienceIn this paper, a transient injection flow at low Mach number regime is investigated. Three different methods are used and analyzed. Two of them are based on asymptotic models of the Navier-Stokes equations valid for small Mach numbers, whereas the other is based on the full compressible Navier-Stokes equations, with particular care given to the discretization at low Mach numbers. Numerical solutions are computed both with or without the gravity force. Finally, the performance of the solvers in terms of CPU-time consumption is investigated, and the sensitivity of the solution to some parameters, which affect CPU time is also performed

Challenges in Containment Thermal Hydraulics

The containment of a nuclear reactor is a component whose loss in an accident has serious consequences on property, persons, and environment. The Fukushima accident reminded us of this reality. For more than 30 years, the French Nuclear Energy and Alternative Energies Commission has been conducting research on the failure modes of these enclosures, particularly on their slow pressurization during a steam release and hydrogen risk. Significant progress has been made on wall condensation and its spatial distribution, the occurrence and erosion of gas stratification, and the impact of mitigation systems, such as spraying and catalytic recombiners. This knowledge has been included in numerical tools and internationally recognized expertise. These tools have also been used for the safety of the hydrogen energy industry. The emergence of new systems, particularly passive systems and new light water reactor concepts, has led us to examine new questions that will have to be addressed in the coming years. This examination is done in view of current computational fluid dynamics code capabilities and limitations

Comparison and analysis of the condensation benchmark results

In the frame of the SARnet Network of Excellence, the need was felt for assessing the status of condensation models adopted in CFD codes relevant for nuclear reactor containment applications. The motivation for this work was provided by the increasingly widespread use of CFD in the analysis of containment behavior consequent to postulated severe accidents, in which wall condensation can promote containment atmosphere mixing. Since standard models are seldom available in many CFD codes for dealing with condensation and in consideration of the different strategies envisaged for analyzing downscaled facilities or full scale containments, this aspect was considered worth of a specific attention. In this aim, after performing a review of the models available to the Participants in the network, appropriate Benchmark Problems were proposed to assess and compare their behavior. The University of Pisa took the charge of coordinating these efforts, proposing an initial step of the Benchmark (identified as the 0th Step) aimed at comparing code responses among each other and with applicable correlations in the application to a classical problem of condensation on a flat plate; the reference geometrical and operating conditions for this step were selected as an idealization of those typical in the CONAN experimental facility, operated at the University of Pisa. Then, the 1st Step of the activity involved addressing experimental data from the CONAN facility at different steam mass fractions and velocities and the comparison of the measured condensation rates and of local heat fluxes with code predictions. Both the steps in this activity were fruitful, since they constituted a gradual and relatively systematic approach to the actual experimental conditions, allowing for revising model details and discussing numerical and physical options. Though the comparison with experiments involved up to now only a limited number of data points, the activity is not considered to be completed and additional experimental data will be offered in the future to obtain a broader assessment of codes in conditions of interest for severe accidents in light water reactors