Tritium Retention and Release Analysis for the U.S.-ITER Driver Blanket

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Abstract
The U.S. design for the ITER tritium-breeding blanket consists of layers of Be multiplier, stainless steel cladding, and Li2O ceramic breeder. Tritium is recovered from the ceramic breeder by purging it with He + 0.2% H2. Models have been developed to describe the purge-flow thermal-hydraulics and gas reactions and the tritium retention/release due to lattice diffusion, desorption/adsorption, solubility/precipitation, and percolation through interconnected porosity. These have been incorporated into the steady-state code TIARA for the purpose of performing design calculations for Tritium Inventory and Release Analysis. Transient calculations for pulsed operation are done with a modified version of the DISPL code. The results of both steady-state and transient analyses for tritium retention and release are given for anticipated ITER operating conditions.