Natural convection test in Phenix reactor and associated CATHARE calculation

• Published in: Nuclear engineering and design Vol. 253; pp. 23 - 31
• Main Authors: Tenchine, D., Pialla, D., Gauthé, P., Vasile, A.
• Format: Journal Article
• Language: English
• Published: Amsterdam Elsevier B.V 01.12.2012; Elsevier
• Subjects: Applied sciences; Controled nuclear fusion plants; Energy; Energy. Thermal use of fuels; Exact sciences and technology; Fission nuclear power plants; Fuels; Installations for energy generation and conversion: thermal and electrical energy; Nuclear fuels; Computational fluid dynamics; Sodium; Fast reactor; Natural convection; Safety; Modeling; Pump; Reactor core; Steam generator; Nuclear reactor
• ISSN: 0029-5493; 1872-759X
• DOI: 10.1016/j.nucengdes.2012.08.001

► Phenix reactor characteristics and instrumentation are briefly described. ► Phenix natural convection test scenario and main test results are presented. ► CATHARE modelling of Phenix primary circuit is depicted. ► Comparison of CATHARE simulation and Phenix data is presented and discussed. The Phenix sodium cooled fast reactor (SFR) started operation in 1973 and was stopped in 2009. Before the reactor was definitively stopped, ultimate tests were performed, including a natural convection test in the primary circuit. One objective of this natural convection test is the validation of plant dynamic codes as CATHARE code for future safety studies on SFRs. The paper firstly describes the Phenix pool type reactor primary circuit with the main components and the instrumentation used during the tests. The initial test conditions and the detailed transient scenario are presented: steam generators dry out, scram, stop of the primary pumps, development of natural convection in the primary circuit with two different phases. Then, CATHARE modelling of the Phenix primary circuit is described. The whole test is calculated for a total duration of 7h in natural convection regime. The CATHARE calculations are compared to the Phenix measurements. A particular attention is paid to the significant decrease of the core power before the scram, due to the increase of temperature at the core inlet. Then, the computed evolution of main components inlet and outlet temperatures is compared to the reactor data. The need of coupling system code with CFD code to model the 3D behaviour of large pools during natural convection regime is pointed out.