Status of TRIO_U code for sodium cooled fast reactors
| Autor: | F. Ducros, Ulrich Bieder, S. Vandroux, C. Fournier, P. Quemere, B. Mathieu, Gauthier Fauchet, D. Tenchine, V. Barthel, F. Perdu |
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| Rok vydání: | 2012 |
| Předmět: |
Nuclear and High Energy Physics
Engineering business.industry Hydraulics Mechanical Engineering Mechanical engineering Nuclear reactor Computational fluid dynamics Plenum space law.invention Thermal hydraulics Nuclear Energy and Engineering Nuclear reactor core law Heat transfer General Materials Science Decay heat Safety Risk Reliability and Quality business Waste Management and Disposal |
| Zdroj: | Nuclear Engineering and Design. 242:307-315 |
| ISSN: | 0029-5493 |
| DOI: | 10.1016/j.nucengdes.2011.10.026 |
| Popis: | Sodium cooled fast reactors (SFRs) have been developed in France for nearly 50 years with successively Rapsodie, Phenix and Superphenix plants. Nowadays, the so-called Astrid prototype is developed in France in the frame of Generation IV deployment. The Astrid project requires thermal hydraulic inputs to support the design and the safety analysis. This paper deals with some thermal hydraulic concerns in the primary circuit: the subassembly, the core, the hot plenum and the cold plenum. The so-called TRIO_U Computational Fluid Dynamic (CFD) code developed at CEA has been progressively adapted to these Astrid concerns. The paper presents the recent improvements, the present status and the remaining challenges for TRIO_U code on each topic. For the subassembly, refined modelling and sub-channel modelling have been developed in parallel. The validation process based on existing experimental data is in progress. A global core modelling including the inter-wrapper region and the connection to the hot plenum is depicted. The need of experimental validation is pointed out. The core outlet region requires refined Large Eddy Simulation computations to predict temperature fluctuations which can induce thermal fatigue. Validation based on sodium experimental data is briefly presented. Thermal stratification in the plenum is a key point for thermal stress analysis on the structures. Validation process includes the comparison to reactor data. Special developments using a Front Tracking method are carried out to deal with free surface and gas entrainment. A methodology including local and global modelling is developed and the validation process is in progress. For decay heat removal situations and especially in natural convection cases, the whole primary vessel – except at the moment the intermediate heat exchangers and the pumps – is modelled with TRIO_U code. Phenix ultimate tests performed in 2009 will be used for the qualification of these particular situations. |
| Databáze: | OpenAIRE |
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