TBM Tritium Processing And Control Dai-Kai Sze UCSD Presented at the US ITER-TBM Meeting Idaho Falls, Idaho Aug. 10-12, 2005 Goal of Tritium Recovery System The TBM tritium recovery system is to demonstrate the tritium recovery system for a DEMO reactor. The tritium recovery system for a DEMO reactor needs to limit the tritium inventory in the blanket system to be lower than 100 g. The tritium recovery system also needs to limit the tritium loss rate to be below 10 Ci/FPD. The cost and the power requirement of the system needs to be below 10 MW, and 10 M$, respectively. Concerns for LiPb blanket The tritium solubility in LiPb is very low. The characteristics of the tritium system are high tritium partial pressure and low tritium inventory. Tritium recovery to a low tritium inventory is not an issue. Tritium recovery to a low tritium partial pressure to satisfy tritium control concern is always the key issue. Literature Review Many tritium recovery processes have been proposed: S. Tanaka “Hydrogen transport through gas/Li17Pb83 interface and effect of isotope swamping” Fusion Technology (1992) 1513T. Terai “Tritium release behavior from molten lithium-lead alloy by permeation through stainless steel type 304, J. Nucl. Mater. 191-194(1992)209 T. Terai “Surface oxide layer as a barrier to tritium permeation through structural material facing 17Li-83Pb molten alloy” Fusion Technol (1992) 1518 A. Viola ”Recent experimental results of the rates of hydrogen permeation through metallic materials in the presence of liquid Pb17Li alloy: influence of the performance and design of tritium recovery equipment” Fusion Eng. Des. 8 (1989) R. Conrad “In-pile release from liquid breeder material Pb-17Li in LIBRETTO 1 experiment” Fusion Technol. (1988) 1046 R. Conrad “Irradiation of liquid breeder material Pb-17Li with in-situ tritium release measurements in the LIBRETTO 2 experiment” J. Nucl. Mater. 179-181 (1991) 875 G. Thevenot “LIPSIE device: Pb-17Li irradiation in water loop with on-line tritium measurement” J. Nucl. Mater. 191-194 (1992) 226 J. Reimann “Tritium separation from fusion LiPb blankets by permeation into a gas stream or into NaK” Fusion Technol. (1986) 1579 G. Pierini “Mass transfer operational units applied to the liquid breeding material of fusion reactors for tritium recovery” Fusion Techoll. 8(1985) 2121 G. Pierini “Tritium recovery from liquid Li17Pb83 alloy blanket materials” Nucl. Eng. Des/Fusion 1 (1984) 159 D. Sze “Counter current exchange extraction system for tritium recovery from 17Li-83Pb” Fusion Technol 8 (1985) 887 G. Pierini “ Feasibility study of tritium recovery systems from 83Pb17Li breeding material of a D-T fusion reactor” (1984) 1429 B. Malara “The feasibility of tritium extraction from blankets of fusion reactors in light of recent experimental data” Fusion Technol., (1992/ 1429 G. Pierini “ The statues of the art of tritium recovery from liquid Pb-17Li blanket materials” Proc. Int. Symp. On fusion reactor blanket and fuel cycle technology, Tokai, Japan, 1986 140-144 F. Feurestein “ Transport of deuterium and rare gases by flowing molten Pb-17Li, J. Nucl. Mater 179-181 (1991) 882 T. Terai “Mass transfer coefficient of tritium from molten lithium-lead alloy (Li17Pb83) to environmental gas under neutron irradiation” Fusion Eng. Design 17 (1991) 272 A Viola “Experimental results on the hydrogen mass transfer rate in molten Pb17Li alloy” Fusion Technol. (1988) 1265 A. M. Polcaro “The kinetics of hydrogen adsorption in molten Pb/Li alloy” J. Nucl. Mater. 119 (1983) 291 J. Riemann “Tritium separation from Pb-17Li by permeation into Na or NaK and cold trapping: First experiments on recovery from cold traps” Fusion Technol. 14 (1988) 804 J. Riemann “ Tritium removal technique for a self-cooled Pb-17Li blanket” 18th SOFT J. Riemann “ Tritium removal from NaK-cold traps: first results on hydride precipitation kinetics” Fusion Technol. 21 (1992) 872 J. Riemann “ Tritium removal from naK-cold traps: investigation of hydrogen release kinetics” Fusion Eng. Des. 18 (1991) 67 H. Feuerstein “Extraction of tritium from Pb-17Li by use of solid getters” Fusion Technol. (1990) 646 M. carolin :The impact of tritium solubility and diffusivity on inventory and permeation in liquid breeder blankets” Fusion Technol 14 (1988) 663 Many works on tritium recovery from LiPb have been published. Most of this proposals can recovery tritium to the inventory limit (<100g). The concern is if any one of those methods can reduce the tritium partial pressure to be below the control limit. Critical assessment of the proposed tritium recovery processes is necessary to decide any of the process can meet the desgin goal. What US should do ITER TBM will not need a tritium recovery process at least until the starting of the D-T phase, or about 17 years from now. We need to concentrate our R/D efforts to the task which is critical to the DCLL concept, such as the SiC or other forms of insert. The US effort will do a through review of the existing tritium recovery literature. We will work with other ITER parties through international collaboration, such as IEA collaboration. We will start tritium R/D, if necessary, maybe 10 years from now.