A verification problem for thermal hydraulics systems codes dealing with twin, parallel-boiling and inverted U-tubes condensing channels

摘要

In the analysis of the nuclear safety of complex nuclear systems, almost one-dimensional system thermal-hydraulics codes will be used perhaps for a couple of decades from now. Computational Fluid Dynamics (CFD) tools are accepted at present to be a support of such analyses and they are used coupled to systems codes or as separate analysis tools for isolated components with boundary conditions obtained from systems codes. The restricted acceptance of "pure" CFD codes is due to many reasons but two of them are relevant, namely (a) the apparent lack of CFD grade experimental data and (b) the need for a complete verification and validation (V&V) and the uncertainty quantification for the codes currently available. There is plenty of experimental data related to integral test facilities (ITFs) that constitute macroscopic systems behavior information and a consolidated data base for such purposes. Despite of this, additional verification cases may be added to the above mentioned consolidated data. In the present paper, flow oscillations in parallel channel configurations with system codes are studied in diverse configurations. Different models, calculation options and, in particular, in-phase or out-of-phase oscillations were studied, both in heated and cooled parallel channels. The emphasis is on the effects of concentrated irreversible pressure losses coefficients at the inlet and at the outlet of the channels. In the case of cooled steam generator channels, the results of the Semiscale Integral Test Facility operating in natural circulation conditions are revisited. The results presented in this paper, show how a validation case lead to find a not still reported (in the Authors knowledge) verification case. The problem is related to twin-parallel-boiling and condensing, inverted U-tubes channels and connected through common plena. This is, of course, a problem that deserved many tens of papers in the last four decades. Flow splitting without reversal was computationally found and to explain this behavior a theoretical model limited in scope was developed that was a posteriori verified using a particular systems code (RELAP5) commonly applied to perform safety analyses of nuclear power plants. The rationale followed, the theoretical analysis performed and the confirmatory computational results found are summarized in this paper.