Page 99 - Thermal Hydraulics Aspects of Liquid Metal Cooled Nuclear Reactors

Rod bundle and pool-type experiments in water serving liquid metal reactors 73

same time as the upper plenum temperature evolution. A separate water model, named
COCO (scale 1/10), was used for the study of the thermal hydraulics in the lower ple-
num. A water model of the Japan Sodium-cooled Fast Reactor (JSFR) at scale 1/10 has
been built to study natural-circulation decay heat removal. The model was built to
address issues in natural-circulation operating conditions identified by performing first
simulationtests.TheRAMONAandNEPTUNwaterfacilities(scale1/20andscale1/5,
respectively) have been developed in Germany to simulate sodium flows. The two dif-
ferent scaling parameters allowed the study of scaling effects. The SAMRAT water
model (1/4 scale of Prototype of Fast Breeder Reactor) was constructed for various
experimental requirements such as thermal-hydraulic studies (including velocity mea-
surements). It was used for the study of free level fluctuations, gas entrainment, thermal
stratification, and thermal stripping (PFBR).
Recently, CEA has identified four water mock-ups in support to ASTRID and their
SFR program (Gu enadou et al., 2015). The MICAS set-up (Fig. 3.1.11A) at 1/6 scale is
dedicated to study flow regime of the internal vessel (hot plenum), both for code val-
idation, engineering design and developments.

Fig. 3.1.11 (A) Sketch of MICAS mock-up from CEA (Gu enadou et al., 2015). (B) Main
elements of a primary loop in a pool-type reactor (left). Natural convection loop (right)
(Spaccapaniccia et al., 2017).