The objective is to simplify Outer Fuel Cycle (OFC) analysis for Fusion reactors through object-oriented description and efficient pre-built functions, in a easy to understand open-source code to do a quick preliminary estimate of design parameters for an attainable OFC. In the current state, the analysis is limited to pipes with a tritiated liquid medium and zero Tritium partial pressure on the outside, which is the equivalent boundary condition for a constant gas sweep or in presence of water on the outside. In the current state, surface effects are implemented only on the outer side of the pipe. This is based on the assumption that inside molten salt fluids the oxyde layer which limits dissociation and recombinarion is not expected to form. In liquid metals Tritium is in the atomic phase, as for the solid metal, so dissociation and recombination are not expected to happen. In its current state, the code is capable enough to simulate heat exchangers, Permeation against vacuum and Breeding blanket, together with their interaction.
To run copy the repo in a local folder, and run:
pip install -r requirements.txt
Contributions are welcome from everyone. The project is a work in progress and may intercur significant and structural changes.
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Fuerst, Thomas F., Chase N. Taylor, and Paul W. Humrickhouse. Tritium Transport Phenomena in Molten-Salt Reactors: Molten Salt Tritium Transport Experiment Design. No. INL/EXT-21-63108-Rev000. Idaho National Lab.(INL), Idaho Falls, ID (United States), 2021.
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