Preliminary Thermal-hydraulic Analysis of Proposed Simplified Test Blanket Module (TBM) as a Subcritical Reactor Driven by Fusion Neutrons

Developing fusion-fission hybrid systems offers a promising pathway to address critical challenges in energy generation and nuclear waste management. This study focuses on the thermal-hydraulic evaluation of a simplified Test Blanket Module (TBM) designed as a subcritical reactor driven by fusion neutrons. The primary objective is to assess the thermal and fluid dynamic performance of the module under steady-state conditions. The TBM is modeled as a subcritical assembly loaded with MOX fuel. Fast neutrons generated by a Reversed Field Pinch (RFP) provide adequate neutron flux to sustain fission reactions within the TBM. In the present study, the fluid dynamics analysis of the molten salt coolant flow, here a mixture of NaF and ZrF ₄ , in a fission blanket for the RFX-mod2 tokamak is carried out. Thermophysical properties of the molten salt composition are considered for temperature conditions between 500 and 900°C. The thermophysical properties of the MOX fuel, which changes in the fuel composition during the blanket operation, are used assuming a burnup equal to 40MWday/kg _HM . Fuel compatibility is well maintained with AISI-316 steel cladding and coolant. To specify its thermal-hydraulic behavior during normal operational conditions, the simulation has been conducted by the ANSYS-Fluent computational fluid dynamics code. Main assumptions and outcomes of this study are presented and critically discussed. The analysis results indicate that the temperature and power distribution within the module are within acceptable limits, and all components demonstrate good compatibility.