Deuterium desorption behavior of solid tritium breeding material, lithium titanate

Several types of blanket module, solid breeder/water or helium cooling, LiPb breeder/helium cooling, liquid LiN and Flibe, have been developed toward both ITER and a demonstration reactor in Japan. In the solid breeder blanket cooled by water, pellets of Li2TiO3 will be employed as tritium breeding material. Structure material in this blanket is low activation ferritic steel, F82H. The operation temperature is limited below approximately 820 K owing to swelling caused by neutron irradiation. Tritium produced by fusion neutrons in this breeding material has to be desorbed under a blanket operation for tritium recovery to be easy. The blanket module, however, has a spatial distribution of temperature. Thus, the tritium desorption behavior has to be clarified in order to make a scheme for tritium recovery. In the present study, a solid breeding material, Li2TiO3, was irradiated by 1.7 keV deuterium ions, and an amount of retained deuterium and deuterium desorption behavior were investigated using a thermal desorption. Dependence of deuterium fluence on amount of retained deuterium was also obtained. In order to examine trapping mechanisms of deuterium in Li2TiO3, similar experiments were conducted for Li and Ti. Deuterium implanted to Li2TiO3 desorbed in forms of HD, D2, HDO and D2O. The amount of deuterium desorbed in form of HD was a few times larger than those of other gas species. The desorption peak appeared at 600 K, but significant desorption up to 900 K was observed. The range of temperature in the lithium titanate of the blanket module is assumed from 550 K to 1200 K. These results suggest that the tritium produced in the blanket is partly not desorbed. Thus, the spatial distribution of temperature in the blanket has to be controlled for the tritium to be desorbed during the operation. The desorption spectra of deuterium in Li2TiO3 were similar to those of Li. This suggests that most of implanted deuterium is trapped in forms of Li-D and Li-OD. Based upon the present results, a tritium inventory and a suitable temperature profile of the blanket module for the inventory to be reduced are discussed. This work is supported by the Grant-in-aid for Scientific Research (No.18360439) of MEXT Japan, and partly Research Collaboration Using Fusion Facilities in JAEA. (authors)