Preliminary nuclear and thermal analysis of a dual-coolant lead-lithium blanket concept for the Helios fusion power plant

This study details the preliminary nuclear and thermal analysis of a dual-coolant lead-lithium (“DCLL”) blanket concept for the “Helios” stellarator fusion power plant. The design employs a vanadium alloy first wall cooled by helium and a tritium breeding zone of flowing lead-lithium eutectic and helium within a EUROFER97 structure. Three-dimensional (“3D”) coupled neutron-photon transport simulations were utilized in OpenMC to verify the design’s key nuclear performance objectives. Results demonstrate a satisfactory idealized Tritium Breeding Ratio (“TBR”) of 1.30 at 65% lithium-6 enrichment. Analysis of high-energy neutron damage predicts a substantial mean first wall lifetime of 17.7 full power years (“FPY”). Radiation shielding criteria from fast neutrons to the superconducting coils were met, with a minimum shaping coil lifetime of 40.3 FPY against a chosen fast neutron fluence limit. The mean nuclear heating in the coils is acceptable at 80.6 W/m 3 , though a small portion of coils exceed the cooling power limit and will require additional cooling or shielding considerations. Thermal analysis performed with AMESim confirms that the design maintains all structural materials within their operational temperature limits, ensuring both high power cycle efficiency and material integrity. This work establishes the basic viability of the DCLL blanket for Helios, providing insight to guide subsequent high-fidelity thermal-hydraulic, magnetohydrodynamic, and neutron transport modeling.