Nuclear Engineering and Design 242 (2012) 307–315 Contents lists available at SciVerse ScienceDirect Nuclear Engineering and Design j ourna l ho me page: www.elsevier.com/locate/nucengdes Status of TRIO U code for sodium cooled fast reactors D. Tenchine ∗ , V. Barthel, U. Bieder, F. Ducros, G. Fauchet, C. Fournier, B. Mathieu, F. Perdu, P. Quemere, S. Vandroux CEA, DEN, DM2S-STMF, F-38054 Grenoble, France a r t i c l e i n f o Article history: Received 3 August 2011 Received in revised form 12 October 2011 Accepted 14 October 2011 a b s t r a c t 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 remain- ing 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 ther- mal 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. © 2011 Elsevier B.V. All rights reserved. 1. Introduction Sodium cooled fast reactors (SFRs) have been developed in France for nearly 50 years with successively Rapsodie, Phenix and Superphenix plants. Phenix reactor started in operation in 1973 and was stooped in 2009. Superphenix reactor went criti- cal in 1985 and was stopped in 1997. The European Fast Reactor project was launched in the late 1980s for several years, in col- laboration with England and Germany. Many thermal hydraulic studies were performed to support design and safety analysis for Superphenix and then for the European Fast Reactor project (Tenchine, 2010). Initially based on analytical and experimental means, thermal hydraulic studies progressively included numerical simulation. In 2006, the decision was taken by the French government to build a first Generation IV prototype by 2020. It was clear that such ∗ Corresponding author. Tel.: +33 4 38 78 30 85; fax: +33 4 38 78 57 28. E-mail address: denis.tenchine@cea.fr (D. Tenchine). a short time schedule could be reached only by designing and build- ing sodium cooled fast reactor with benefit of the past experience. The prototype is called Astrid. New thermal hydraulic challenges are concerned with the Astrid project to improve safety and econ- omy of the future Generation IV sodium cooled fast reactors. The paper presents selected applications of the TRIO U Com- putational Fluid Dynamic (CFD) code, performed within the Astrid project, as well as selected validation test cases. TRIO U has been developed at CEA for many years and it is used for numerous appli- cations such as various Pressurised Water Reactors, nuclear fuel cycle and other industrial items. The paper presents the recent developments, the present status and the remaining challenges for TRIO U code on each of the following topics: - For the subassembly, refined modelling and sub-channel mod- elling 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 pool is depicted. The need of experi- mental validation is pointed out. 0029-5493/$ – see front matter © 2011 Elsevier B.V. All rights reserved. doi:10.1016/j.nucengdes.2011.10.026