Deuterium retention and desorption behavior of lithium titanate

In a solid 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 assumed to be as high as 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 known 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 approximately one order of magnitude larger than those of other gas species. The desorption peak appeared at 600 K, but significant desorption up to 900 K was observed. The temperature range in the blanket is assumed from 550 K to 1200 K. These results suggest that the tritium produced in the blanket is partly not desorbed. Thus, the temperature spatial distribution 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 form of Li-D and Li-OD. Based upon the present results, suitable design of blanket components is 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. (orig.)