A Study on flow field of purge gas for tritium transfer through breeder pebble bed in fusion blanket

Seki, Yohji; Ezato, Koichiro; Yokoyama, Kenji; Enoeda, Mikio; Kubota, Jinichi*; Sakamoto, Kensaku 

Japan Atomic Energy Agency has been performing R&D and design of a blanket module of a nuclear fusion reactor. Pebbles of a ceramic tritium breeder are packed in a container of the blanket. Helium purge gas is applied as a transport fluid in a tritium recovery system. Prediction of the flow phenomena with a tritium transfer is important for designs of the container. A purpose of our research is to establish and verify a method for a prediction of the flow in the pebble bed. In this study, pressure drops of the helium purge gas through the pebble bed were measured up to 100 L/min of flow rate. Reliability of prediction ability of the pressure drop was validated by this experiment within the flow rate which is less than 40 L/min. A numerical simulation for the flow field through the pebble bed also has been performed. Consequently, the velocity distributions are quantitatively and qualitatively obtained at near the wall and the center region in the pebble bed.