Activity assessment in water coolant loops and rooms using the SAETTA tritium transport code

The management of hydrogen isotopes within a tokamak fusion reactor presents a significant design challenge, particularly with respect to tritium management. It is essential for the fusion power plant to achieve self-sufficiency in its fuel supply while minimizing component contamination and external releases. This study presents an analysis of tritium migration within the EU DEMO Water-Cooled Lead–Lithium (WCLL) Breeding Blanket (BB) Primary Heat Transfer System (PHTS). The objective is to evaluate tritium inventories and release rates from the primary circuit to the tokamak building. The assessment utilizes SAETTA, a validated system-level tritium transport code, able to describe several physical phenomena in a variety of medium and structural components. A versatile PHTS model has been developed, allowing for relatively rapid yet detailed sensitivity analyses on specific parameters. The tritium inventory in the primary coolant and connected rooms has been estimated, along with loss rates. Key factors impacting inventories and losses under reference conditions include the size of the Coolant Purification System (CPS) and potential leaks from piping. Additionally, the study discusses the acceptable limits for tritium concentration in the air. The effects of leaks, CPS size, hydrogen concentration in the water, and Permeation Reduction Factors (PRF) at the steam generator walls have also been evaluated. The findings from this analysis will aid in optimizing the PHTS circuit design, defining tritium management and room ventilation strategies, and informing the design of fuel cycle components by providing necessary source terms.