Technical Meeting on Tritium Breeding Blankets and Associated Neutronics

Performance Assessment and Integration of Different Tritium Extraction Technologies in the Outer Fuel Cycle

2 Sept 2025, 12:10

30m

IAEA Headquarters, Vienna, Austria

Oral Track I: Breeding blanket design and performance Topic I

Speaker

Gabriele Ferrero (Politecnico di Torino)

Description

The outer fuel cycle (OFC) of a fusion power plant (FPP) is a fundamental component that allows retrieving the bred tritium to permit D-T plasma operations. Moreover, this system plays a fundamental role for the safety of workers and population, acting as a physical barrier that delimits the tritiated circuit, and extracts tritium from the system, thereby diminishing the tritium inventory both in the BB and in ancillary systems.
The open-source TRItium Object-oriented and Modular Analysis (TRIOMA) code is a Python package the analysis of OFCs in FPPs. The code leverages on pre-built classes and functions to build an intuitive, object-oriented framework. Its 0-dimensional analytical formulation enables fast calculation of the extraction efficiency and of integral parameters, including tritium inventory and losses, under steady-state conditions, for each component of the OFC, enabling rapid design iterations for OFC optimization. Both molten salt and liquid metal breeders, which have different analytical formulations due to different solubility laws [1], [2], [3], are included in TRIOMA. The code can be easily employed to make the preliminary design of an OFC, to analyze feasibility and sensitivity on input properties, to observe start-up transients, and pulsed plasma operations. Moreover, due to the flexibility of the Python environment, the integration with other open-source tritium transport tools and data analysis is facilitated.
The presented work employs TRIOMA to design and compare different tritium extraction technologies within the same OFC circuit. In particular, the comparison involves the Permeator Against Vacuum (PAV), the Gas Liquid Contactor (GLC) based on a packed tower column, and the Liquid Vacuum Contactor (LVC) technologies. These are the main selected tritium extraction technologies for PbLi liquid breeders foreseen in the WCLL breeding blanket module. In particular, the PAV mock-up experiment is based at ENEA Brasimone [4] in the TRIEX facility [5], as well as the GLC mock-up, while the LVC technology is tested in CIEMAT laboratories.
REFERENCES
[1] P. W. Humrickhouse and B. J. Merrill, “Vacuum Permeator Analysis for Extraction of Tritium from DCLL Blankets,” Fusion Science and Technology, vol. 68, no. 2, pp. 295–302, Sep. 2015, doi: 10.13182/FST14-941.
[2] J. D. Rader, M. S. Greenwood, and P. W. Humrickhouse, “Verification of Modelica-Based Models with Analytical Solutions for Tritium Diffusion,” Nuclear Technology, vol. 203, no. 1, pp. 58–65, Jul. 2018, doi: 10.1080/00295450.2018.1431505.
[3] C. Alberghi, L. Candido, M. Utili, and M. Zucchetti, “Development of new analytical tools for tritium transport modelling,” Fusion Engineering and Design, vol. 177, p. 113083, Apr. 2022, doi: 10.1016/j.fusengdes.2022.113083.
[4] F. Papa et al., “Engineering design of a Permeator Against Vacuum mock-up with niobium membrane,” Fusion Engineering and Design, vol. 166, p. 112313, 2021.
[5] M. Utili, C. Alberghi, L. Candido, F. Papa, M. Tarantino, and A. Venturini, “TRIEX-II: an experimental facility for the characterization of the tritium extraction unit of the WCLL blanket of ITER and DEMO fusion reactors,” Nucl. Fusion, vol. 62, no. 6, p. 066036, Apr. 2022, doi: 10.1088/1741-4326/ac5c74.