Recent advances in tritium modeling and its implications on tritium management for outer fuel cycle

Abstract Recent advances in tritium transport modeling of helium-cooled ceramic breeding blankets systems has shined light into some tritium management issues. A detailed component model accounting for multi-physics, design, and operational features is necessary to provide accurate estimations of tritium permeation rates to the building/environment- a safety and licensing concern for a fusion nuclear reactor. We found that tritium permeation to buildings can be reduced of ∼20 times when H2 is increased from ∼0.2 Pa to 100 Pa in coolant streams due to the effect of H and T co-permeation. Similarly, the practice of adding about 0.1 % vol of H2 into the helium purge gas to promote tritium release can also reduce permeation from breeding zones to coolant systems. However, high H2 partial pressure in helium purge gas further complicates tritium extraction methodology, and may compromise extraction efficiency. This paper provides a concentrated analysis of tritium management in the He-cooled ceramic blankets with the goal of providing further outer fuel cycle tritium R&D guidance from an integrated point of view.