Assessment of the Activation, Decay Heat, and Waste Disposal of a Dual Coolant Lithium Lead Test Blanket Module for ITER

摘要

The US is proposing a test blanket module (TBM) to be placed in half of the three dedicated test ports of ITER. The TBM is based on the dual coolant lithium lead (DCLL) blanket concept. Conventional ferritic steel (F82H) is used as the structure of the first wall (FW), the two breeder channels, the back plate, the inlet/out piping, and the shield plug. Two separate cooling circuits are employed: helium is used to cool the FW and blanket structure while the Pb-17Li is used as a coolant and breeder mainly in the two breeder channels. SiC flow channel inserts (FCI) are used to thermally and electrically isolate the flowing Pb-17Li from the relatively low-temperature structure. A 2 mm thick beryllium layer is used as a plasma facing material on the FW area (1.25 m2 ) subjected to 0.78 MW/m2 neutron wall load. In this paper, we present results pertaining to the radioactive inventory and decay heat levels at shutdown and at several post-irradiation times following the pulsed operation scheme of ITER