Analysis of the Natural Convection Flow in the Upper Plenum of the MONJU Reactor with Trio_U

The IAEA has coordinated a benchmark project on natural convection phenomena in the upper plenum of the MONJU reactor. JAEA has provided both detailed geometrical data of the plant and complete thermalhydraulic boundary conditions describing a pump trip transient, accomplished during the start-up experiments of the reactor. For the initial conditions of the pump trip transient, extensive sensitivity analyses have been made with the CFD code Trio_U. These calculations show a high sensitivity of the global flow pattern in the MONJU upper plenum depending on the initial order of the numerical scheme and the modelling of the geometrically complex upper core structure. During the pump trip, the formation of a thermal stratification within the plenum has been observed which persists for almost two hours. All calculations have shown a homogenization of the temperature in the plenum after about 15 minutes. A slight reduction of the mixing in the upper plenum could have been achieved by modifying the form of the flow holes in the inner barrel (fillets instead of sharp edges) in order to reduce their axial pressure loss. 1. Introduction The IAEA has coordinated a research project (CRP) between 2008 and 2012 entitled “Benchmark Analysis of sodium Natural Convection in the upper plenum of the MONJU Reactor Vessel.” Eight research organizations from seven countries with an active program on sodium cooled fast reactors—namely, China, France, India, Japan, Republic of Korea, Russian Federation, and USA—contributed to this CRP. Japan Atomic Energy Agency (JAEA) has submitted to the CRP participants the data of sodium thermal stratification measurements in the MONJU reactor vessel upper plenum collected during a plant trip test conducted in December 1995. The benchmark partners have analysed this experiment by applying different codes and methodologies. The benchmark thus helped the members to improve their capability in the field of fast reactor in-vessel Sodium thermalhydraulics. Sodium cooled fast breeder reactors are under development for more than 50 years. Nevertheless, only very limited data are published to date which allow the validation of CFD codes in general and for natural and mixed convection phenomena in particular. Within the European Fast Breeder Reactor project, an experimental approach in the RAMONA facility has been largely used to study decay heat removal situations [1]. RAMONA was a 1/20 scale water model of the upper plenum with boundary conditions imposed at the core outlet and with active immersed coolers. Transient situations were tested and