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Description
In the Divertor Tokamak Test (DTT) facility, the Limiter Inboard First Wall (LIFW) plays a key role in shielding the standard inboard wall modules from direct plasma contact during plasma-limited operation, such as ramp-up and ramp-down phases, negative triangularity configurations, transient events, and accidental scenarios. The limiter design relies on modules that protrude radially toward the plasma and feature a toroidally contoured plasma-facing surface, with the objective of enhancing the spatial distribution of heat loads during wall interaction [1]. Although this design approach is expected to reduce localized thermal peaks, the definition of the most effective limiter shaping is still ongoing and calls for a quantitative evaluation of plasma-driven heat fluxes under realistic DTT operating conditions.
In this work, heat load simulations of the DTT Inner Limiter are performed using the PFCflux code to support the ongoing optimization of the LIFW geometry and to quantify the thermal loads associated with representative plasma scenarios. Magnetic equilibria corresponding to plasma-limited configurations are used as input to model the power flow in the scrape-off layer and its interaction with the limiter surface [2]. The resulting spatial and temporal distributions of heat flux are analyzed for different limiter shapes and plasma positions, with the aim of identifying configurations that minimize peak heat loads while preserving effective plasma protection.
The analysis shows that the limiter geometry, including toroidal shaping and radial protrusion, affects the spatial distribution of heat loads on the limiter surface. The predicted heat fluxes also exhibit sensitivity to the assumed plasma configurations, indicating that both geometric and operational aspects should be considered in the limiter design assessment.
This study represents a continuation of the limiter design activity in DTT, bridging preliminary geometric concepts and engineering validation. By integrating plasma heat load simulations into the design workflow, it contributes to the definition of a reliable Inner Limiter configuration capable of withstanding the demanding operational scenarios foreseen for DTT and future fusion devices [3].
- De Luca R., et al. (2025). Thermo-structural assessment of the limiter inboard first wall design of the Divertor Tokamak Test facility. Fusion Engineering and Design, 219, 115281.
- Firdaouss, M., et al. (2013). Modelling of power deposition on the JET ITER like wall using the code PFCFLux. Journal of Nuclear Materials, 438, S536-S539.
- Romanelli, F.,et al. (2024). Divertor Tokamak Test facility project: status of design and implementation. Nuclear Fusion, 64(11), 112015.